ML20155C958
| ML20155C958 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/29/1998 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18039A590 | List: |
| References | |
| RTR-REGGD-01.150, RTR-REGGD-1.150 ADM-BFN-1006V, ADM-BFN-1006V0, NUDOCS 9811030069 | |
| Download: ML20155C958 (46) | |
Text
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O GE Nuclear Energy PROCEDURE: ADM-BFN-1006VO REVislON No.: 0 TITLE PROCEDURE FOR COMPLIANCE WITH USNRC REGULATORY GUIDE 1.150 THIS PROCEDURE IS APPROVED FOR USE AT BROWNS FERRY NUCLEAR AND IS CONTAINED IN GE NDE MANUAL 9.6 OF TVA'S DOCUMENT CONTROL SYSTEM
)
Disclaimer of Responsibility This document was prepared by the General Electric Company (GE). Neither GE nor any of the i
contributors to this document:
A.
Makes any warranty or representations, express or implied, with respect to the accuracy, completness, or usefulness of the information contained in this document, or that the use of any information disclosed in this document may not infringe privately owned rights; or B.
Assumes any responsibility for liability or damage of any kind which may result from the use of any information disclosed in this document 9811030069 981029 PDR ADOCK 05000296 C-P PDR
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No: ADM-BFN-1006VO Revision No.: 0 Page 1 of 6
Title:
PROCEDURE FOR COMPLIANCE WITH USNRC REGULATORY GUIDE 1.150 GE Nuclear Energy TABLE OF CONTENTS I
SECTION DESCRIPTION PAGE NO.
1.0 SCOPE 2
2.0 REFERENCES
2 3.0 METHOD (S) OF 2
COMPLIANCE 4.0 PERSONNEL 6
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No: ADM-BFN-1006VO Revision No.: 0 Page 2 of 6
Title:
PROCEDURE FOR COMPLIANCE WITH USNRC REGULATORY j
GUIDE 1.150 GE Nuclear Energy 1.0 SCOPE 1.1 The equipment certification methods described herein are applicable to both remote automatic and manual ultrasonic equipment used to perform inservice examinations of
" Reactor Vessel Welds" in accordance with a TVA approved GE specific examination procedure.
Quality Assurance requirements are specified in the TVA approved GE specific examiniation procedure which references this administrative document.
1.2 In this procedure, " Reactor Vessel Welds" are those assembly welds that are categorized as either B-A or B-D under the rules of ASME Section XI.
1.3 The purpose of this procedure is to assure that activities described herein are performed and that records specified by Reg. Guide 1.150 are generated, gathered, stored, and reported in a manner consistent with both Reg. Guide and the Owner's Quality Assurance requirements.
1.4 It is not the purpose of this procedure to provide definitive instructions to the operator of the analysis and recording equipment used to generate the required records.
It is, however, a requirement of this procedure that the analysis and recording equipment used be operated in accordance with the manufacturer's recommendations.
2.0 REFERENCES
2.1 Codes and Standards. The following Codes and Standards form a part of this procedure to the extent specified herein.
2.1.1 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
2.1.1.1 Section V " Nondestructive Examination" 1974 Edition, Summer 1975 Addenda, through the latest USNRC approved edition.
2.1.1.2 Section XI " Inservice Inspection of Nuclear Power Plant Components" 1974 Edition Summer 1975 Addenda through the latest USNRC approved edition.
2.1.2 United States Nuclear Regulatory Commission (USNRC).
2.1.2.1 Regulatory Guide 1.150, Revision 1, February 1983, " Ultrasonic Testing of Reactor Vessel Welds During Pre-service and Inservice Examination"
- alternate method.
Note:
In this procedure, * (asterisk) corresponds with the section numbers of the Revision 1 alternate method of Regulatory Guide 1.150.
3.0 METHOD (S) OF COMPLlANCE Note:
The method (s) of compliance with Reg. Guine 1.150 are delineated in the subparagraphs that follow. Each subparagraph corresponds with a section of Reg. Guide 1.150 alternate method. Note that Sections 4 and 5 of Reg.
Guide 1.150 have been deleted from the Revision 1 alternate method.
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No: ADM-BFN-100SVO Revision No.: 0 Page 3 of 6 j
Title:
PROCEDURE FOR COMPLIANCE WITH USNRC REGULATORY GUIDE 1.150 GE Nuclear Energy 3.1 Inspection System Performance Checks 3.1.1 Pre-Exam Performance Checks specified in Paragraph *1.1 will not be performed separately. These checks are identical to the checks required in 3.1.2 below.
3.1.2 Field Performance Checks specified in Paragraph *1.2 shall be performed before, during and after the examination of " Reactor Vessel Welds" as specified below.
3.1.2.1 RF Waveform and frequency amplitude information shall be recorded for each transducer or transducer / wedge combination involved with examination.
These records may be photographic, computer generated, or a combination of both. The reflector used to generate the required RF return signal, as well as the electronic system used, shall be documented. This allows future RF Waveform data to be gathered from the same or a similar reflector, j
3.1.2.2 Screen Height Linearity data shall be gathered in accordance with the requirements specified in the procedure being used. The record of screen height linearity shall be the Calibration Data Sheet.
3.1.2.3 Amplitude Control Linearity data shall be gatherd in accordance with the requirements specified in the procedure being used. The record of amplitude controllinearity shall be the Calibration Data Sheet.
3.1.2.4 Angle Beam Profile Characterization shall be performed for each transducer / wedge combination used to examine " Reactor Vessel Welds" Beam profiling shall be performed both before the search unit is used to examine the first weld and after the search unit is used to examine the last weld during the examination. The data shall be gathered in accordance with requirements of the applicable examination procedure. The record of vertical beam profiles shall be per the requirements of the procedure (s) used during the examination.
3.1.2.5 Transducers and wedges used to examine reactor vessel welds are subject to failures such as connector damage, wedge cracking, and internal electrical malfunctions.
When a failure renders a transducer or transducer / wedge combination inoperable, the post-exam performance checks required in 3.1.2.1 and 3.1.2.4 cannot be performed.
3.1.2.6 The intent of pre-and post-exam checks is to assure there was no gross degradation of performance during a period of use. When post-exam checks cannot be performed, a review of calibration and examination data gathered with the inoperable equipment is required.
If the review shows that performance during the period of the failed equipment's use was satisfactory, the intent of post-exam checks has been satisfied.
3.1.2.7 Documentation that the review was performed fulfills the post-exam requirement for failed equipment.
3.2 Calibration 3.2.1 Calibration for Manual Examinations shall be performed in accordance with the applicable manual ultrasonic examination procedure. Calibration checks shall be
O No: ADM-BFN-1006V0 Revision No.: O Page 4 of 6
Title:
PROCEDURE FOR COM,P LANCE WITH USNRC REGULATORY e
GUlDE 1.150 GE Nuclear Energy performed at the intervals specified in the procedure.
The record of manual ultrasonic calibrations shall be the Calibration Data Sheet.
3.2.2 Calibration for Mechanized Scanning shall be performed in accordance with the applicable remote ultrasonic examination procedure. Calibration checks shall be performed at the intervals specified in the procedure.
The record of remote ultrasonic calibrations shall be the Calibration Data Sheet.
3.2.3 Calibration Confirrnation shall be as specified in the applicable ultrasonic examination procedure (Ref. 3.2.1 - 3.2.2).
3.2.4 Calibration Blocks used during examination of Reactor Vessel Welds shall be those furnished by the Owner. Where possible, the same calibration block (s) should be used to perform successive RPV examinations. Where use of the same block (s) is not possible, the provisions of Section *2.4 shall apply.
3.3 Examinations 3.3.1 The scope and extent of examinations shall be in accordance with Reference 2.1.1.2.
The use of electronic gating is addressed in the applicable ultrasonic examination procedure (s).
3.3.2 Internal Surface examination requirements per Section *3.1 are demonstrated j
during each angle beam calibration. The demonstration documents the ability of the ultrasonic examination system to differentiate between an indication from " clad roll" and an indication from the ID notch in the caiibration standard. The record of this demonstration shall be the Calibration Data Sheet.
3.3.3 Scanning Weld-Metal Interface The ultrasonic examination procedures developed by GE for inservice inspection of the RPV utilize the ASME Section V, Article 4, specified beaming angles of 0*, 45" and 60' These search unit angles are not based on weld preparation geometry; therefore, the ability of these angles to detect a planar flaw at the clad / base metal interface was verified during procedure development.
Documentation is to be included in the " Reporting of Results" required by Paragraph 3.5.
3.4 Recording and Sizing Note:
The capability of GE examination procedures to detect and size flaws during inservice inspections was verified during procedure development.
Documentation is to be included in the " Reporting of Results" required by Paragraph 3.5.
The procedures developed by GE for inservice inspection of Reactor Vessel Welds require all data to be recorded to 20% DAC end points and, if applicable, 50% DAC,100% DAC, and one-half maximum amplitude end points.
All data is recorded at scanning intervals of 1/4" or less.
Determination of the adequacy of the recorded data ord data evaluation
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No: ADM-BFN-1006VO Revision No.: 0 Page 5 of 6
Title:
PROCEDURE FOR COMPLIANCE WITH USNRC REGULATORY GUIDE 1.150 GE Nuclear Energy must be made by a certified Level lli individual other than the one performing the examination.
3.4.1 Geometric Indications: All indications will be evaluated by a certified Level 111. If Level 111 analysis of the data shows that the indication is geomeiric in nature, the basis for the determination will be described in the report of the ;< amination.
3.4.2 Indications With Chanaina Metal Path:
All indications will be evaluated by a certified Level lli individual. The Level lil will determine the location, through-wall dimension, and proper DAC evaluation. No determinations are to be made by the examination teams since, by procedural requirements, all data shall be rrcorded
, iiit equals or exceeds 20% DAC.
3.*.3 Indications Without Chanaina Metal Path: See explanation in 3.4.2 above. Also, see precautionary note in *6.3 C for consideration.
j 3.4.4 Evaluation of Indications:
All indications without changing metal path and i
indications in the outer 75% of the RPV wall thickness shall be sized using 50%
DAC end point data. All indications with changing metal path in the inner 25% of the RPV wall thickness shall be sized using both 20% DAC and 50% end point data.
The size of the indications shall be the larger of:
a)
Indication size determined using 50% DAC end point data,
-OR-b)
Indication size determined using 20% DAC end point data, corrected for the j
predetermined 20% DAC beam spread.
i 3.5 Reporting of Results Note:
The records detailed in Section *7 become a part of the report of the examination. Retention of these records is the responsibility of the Owner.
Any flaw that exceeds the allowable limits detailed in Section XI, will be reported to the Owner in accordance with contractual requirements. The reporting of any flaw (s) indicative of " Abnormal Degradation of Reactor Pressure Boundary"is the responsibility of the Owner.
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[*['3 No: ADM-BFN-1006VO Revision No.: 0 Page 6 of 6
(
Title:
PROCEDURE FOR COMPLIANCE WITH USNRC REGULATORY GUIDE 1.150 GE Nuclear Energy 3.5.1 The best estimate of the tolerances (error band) in sizing flaws has been determined using standards developed by GE for this purpose. The standards for manual equipment and automated equipment represent the range of thicknesses to be examined on the RPV. The data and engineering analysis used to develop these error bands is retained by GE. The determined error band is a part of the report documenting the requirements of Section *7.
3.5.2 The effectiveness of the ultrasonic examination procedures was demonstrated on the standards used for error band determinations (3.5.1 above). Documentation of these demonstrations is a part of the Section *7 report.
3.5.3 The actual volumes that have not been examined due to vessel configuration, such as nozzle interferences or volumes that are shadowed by vessel flaws, must be documented in the report of the examination. Volumes not effectively examined due to near field effects, cladding to base material interface, electronic gating, or opposite surface interferences was determined using the standards discussed in Paragraph 3.5.1.
The determined effects are a part of the Section *7 report. The calculated effect of these items is applied to the results of inservice examinations.
Areas not examined (configuration) or not effectively examined (near field, etc.) will be reported to the Owner in accordance with contractual requirements. Any " Relief Requests" necessitated by these items are the Owner's responsibility.
3.5.4 Sketches and calculations documenting the effect of vessel configuration on inservice examinations of the RPV must be included in the report of the examination.
3.5.5 Sketches of mechanized scanning equipment with necessary reference points and dimensions are a part of the Section *7 report. These sketches allow reviewers to follow the mechanized scanning equipment's indication location method.
3.5.6 Alternative examination techniques, if used, will be documented and the results of such examination will be included in the report of the examination.
4.0 PERSONNEL 4.1 Ail calculations and reporting required by Reg. Guide 1.150 will be performed by a certified Level lli individual. Data required to perform these calculations and evaluations may be gathered by either a certified Level ll or Level ill individual.
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ATTACHMENT E i-i i
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. BFN Unit 3 Inservice Inspection Calibration Standards For the'Second Ten-Year Inservice Inspection Interval i
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BROWN FERRY NUCLEAR PLANT INSERVICE INSPECTION CAllBRATION STANDARDS FOR UNIT 3 SECOND TEN l
YEAR ISIINSPECTION INTERVAL CALIBRATION GENERAL WALL MATERIAL TYPEI COMMENTS BLOCK NO.
DESCRIPTION THICKNESS DESCRIPTION BF-1 4" SCH 80 0.326" A-106 GRADE B CS PIPING BF-2 4" SCH 80 0.326" A-358 TP-304 SS PIPING BF-3 10" SCH 40 0.375" A-106 GRADE B CS PIPING BF-4 6" SCH 140 0.752" A-358 TP-304 SS PIPING BF-5 10" SCH 100 0.730" A-106 GRADE B CS PIPING BF-18 24"X 6"X 6" FLAT BLOCK 6.125" A-533 GRADE B CS RPV NOZZLE AND SHELL WELDS BF-19 24" X 6" X 4" FLAT BLOCK 4 1/64" SQ-533 GRADE B CS RPV HEAD & NOZZLE WELDS BF-26 12" X 4" FLAT BLOCK 0.75*
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BF-29 4" SCH 160 0.506" A-106 GRADE B CS PIPING BF-30 6" SCH 160 0.754" A-106 GRADE B CS PIPING BF-31 6" SCH 80 0.422" 316L SS PlPING BF-33 20" X 4" X t" FLAT BLOCK 1.272" 304L SS PIPING
)
BF-34 6" SCH 80 0.399*
ASTM A-53 GRADE B CS PIPING BF-35 8" SCH 100 0.520" A-106 CS PIPING
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BF-37 14" SCH 30 0.364" SA-106 GRADE B CS PIPING BF-41 4' SCH 159 0.452" SA-376 TP-304 SS PIPING BF-42 10" SCH 100 0.697" SA-376 TP-304 SS PIPING i
BF-46 10" SCH 80 0.623" A-312 TP-304 SS PIPING BF-56 14" SCH 100 0.938" A-106 GRADE B CS PIPING BF-57 20" X 5" X 1" FLAT BLOCK 1.269" 304L SS PIPING BF-59 5" DIA. PIPE SEGMENT 0.923" SA-182 F-316 SS PIPING BF-60 5" SCH XXS 0.758" SA-312 T-316L SS PIPING BF-65 12" SCH 160 1.345" SA-312 T-304 SS PIPING BF-70 13.75" PIPE SEGMENT 1.3" SA-376 TP-316 SS PIPING BF-72 CLAD BLOCK FOR RPV NOZZLES 1.320" ASTM A-508 CLASS 2 CS RPV NOZZLES I
f BF-73 RPV STUD BLOCK 6" DIA.
ASTM A-540 RPV HEAD STUDS BF-75 12" SCH 80 0.688" SA-106 GRADE B CS PIPING BF-76 5.38" OD X 12" CLAD BLOCK 0.846" SA-508 CL 2 CS PIPING / NOZZLE i
BF-77 16" SCH 30 0.375" SA-106 GRADE B CS PIPING BF-79 12" SCH 80 0.691" SA-358 GRADE 304 SS PIPING BF-81 RPV HEAD NOZZLE INNER RAD VARIES SA-508 CL 2 CS RPV HEAD NOZZLE INNER RADIUS BF-82 28" DIA. PlPE SEGMENT 1.250" ASTM A-358 GRADE 304 SS PIPING BF-83 28" DIA. PIPE SEGMENT 1.750" A-358 GRADE 304 SS/308L PIPING I
BF-84 N1, N2, & N3, NOZZLE IR VARIES SA-508 CL 2 CS RPV NOZZLE IR"S L
l BF-85 N2, N5, & N9 NOZZLE IR VARIES SA-508 CL 2 CS RPV NOZZLE IR"S l
i Page 1 f
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BROWN FERRY NUCLEAR PLANT INSERVICE INSPECTION CAllBRATION STANDARDS FOR UNIT 3 SECOND TEN YEAR ISIINSPECTION INTERVAL i
j CAUBRATION GENERAL WALL MATERIAL TYPE /
I COMMENTS BLOCK NO.
DESCRIPTION THICKNESS DESCRIPTION
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BF-86 N8 NO771 F 1R CLAD VARIES SA-508 CL 2 CS RPV NO771 F IR BF-87 20" SCH 80 1.031" SA-376 TP-316 SS PIPING BF-88 22" PIPE 1.031" SA-376 TP-316 SS PIPING BF-89 24" PIPE 1.285" SA-376 TP-316 SS PIPING j
BF-90 28* PIPE 1.44" SA-376 TP-316 SS PIPING BF-91 6" SCH 120 0.562" SA-106 GRADE B CS PIPING BF-92 S* SCH 40 0.322" SA-106 GRADE B CS PIPING I
BF-93 12" SCH 100 0.b43" SA-106 GRADE B CS PIPING i
BF-94 16" SCH 100 1.031" SA-106 GRADE B CS PIPING i
BF-95 18" PIPE 0.312" SA-106 GRADE B CS PIPING
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BF-96 18" PIPE 0.500" 3A-106 GRADE B CS PlPING BF-97 18" SCH 80 0.938" SA-106 GRADE B CS PlPING i
BF-98 20" PIPE 0.500" SA-106 GRADE B CS PIPING i
BF-99 20" SCH 100 1.281*
SA-106 GRADE B CS PIPING BF-100 24" PIPE 0.375 SA-106 GRADF R CS PIPING BF-101 24" PIPE 0.500*
SA-106 GRADE 8 CS PIPING BF-102 24" SCH 100 1.531" SA-106 GRADE B CS PIPING j
BF-103 24" PIPE 2.50P SA-106 GRADE B CS PIPING 4
BF-104 26" PIPE 0.950" SA-106 GRADE B CS PIPING
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BF-105 30" PIPE 0.375 SA-106 GRADE B CS PIPING 4
BF-106 10" SPRAY NOZZLE 1.358 SA-508 CL 2 CS SAFE END TO RPV NOZZLE BF-107 REClRC. OUTLET NOZZLE 1.737" SA-508 CL 2 CS SAFE END TO RPV NOZZLE BF-108 24" PIPE 1.218" SA-106 GRADE B CS PIPING l
BF-109 6.6" VESSEL BLOCK 6.6" SA-533 GRADE B CS REACTOR PRESSURE VESSEL BF-119 3" DIA. RECIRC. PUMP STUD 3" DIA.
A-540 RECIRC. PUMP STUDS
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BF-126 RPV STUD BLOCK 6.75" DIA.
SA-540 GRADE B23 RPV HEAD STUDS BF-127 12" STD SCH 0.390" SA-106 GRADE B CS PIPING BF-128 20* STD SCH 0.375" SA-106 GRADE B CS PIPING BF-129 RPV INST. NOZ. TO FLANGE 7-1/2" OD 1.0" ASTM A-105 GRADE B CS RPV HEAD NOZZLES BF-130 RPV INST. NOZ. TO FLANGE 5 -trE OD 1.0" ASTM A-105 GRADE B CS RPV HEAD NOZZLES BNP-17 12" SCH 40 0.380" A-312 TP-304 SS PlPING SQ-37 12" X 4" X 5" FLAT BLOCK 0.502" A-36 CS PIPING SQ-85 2" SCH 80 0.218" SA-376 TP 304 SS RPV INTSTR. SAFE ENDS f
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Page 2 f
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l ATTACHMENT F Statement Verifying the Personnel Certification Program Complies With the Requirements of.ASME Code Section XI, 1989 Edition, Paragraph IWA 2300, Including Appendix VII l
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To: H. E. Hodges
Subject:
Browns Ferry Unit 3 ISI Second Ten Year Interval-NRC RAI Item F TVA procedure IEP-200," Qualification and Certification Requirements for TVA Nuclear (TVAN) Nondestructive Examination (NDE) Personnel" was revised to meet the requirements of ASME Code Section XI,1989 Edition, paragraph IWA 2300, including Appendix VII requirements for f
ultrasonic personnel. Revision I was approved on May 16,1995, and implemented on August 16,1995.
j TVA Inspection Services Organization's (ISO) NDE certification program was audited by the Hartford Steam Boiler Inspection and Insurance Company on August 11,1995, for compliance to SNT-TC-I A, 1984 Edition ASME Code Section XI,1989 Edition, paragraph IWA 2300 including Appendix VII.
The program was acceptable as meeting the requirements listed above with comments which were l
included in subsequent revisions or answered satisfactorily.
f Contractors on TVA's Approved Suppliers List ( ASL) approved for performing ISI NDE examinations also have revised their programs accordingly to meet these requirements.
I trust this will satisfy the concerns about the NDE qualification program utilized for the Browns Ferry Unit 3 ISI program. Should you have any further questions or need additional information, please contact me at 423-843-4138.
T. L. Stocklem NDE Specialist
2 ATTACHMENT G
1.
Relief Request 3-ISI-3, Revision 1, Longitudinal Welds in Class 1 Piping, with ASME Code Case N-524,
" Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Section XI, Division 1" i
- 2. Relief Request 3-ISI-6, Revision 1,
" Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000;Section XI, Division 1," ASME Code Case N-532 1
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I TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT
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UNIT 3 ASME SECTION XI, INSERVICE INSPECTION PROGRAM, SECOND TEN-YEAR INSPECTION INTERVAL REQUEST FOR RELIEF 3-ISI-3, Revision 1 Executive Summary:
Pursuant to 10 CFR 50.55a (a) (3) (i), TVA is requesting relief from the ASME Section XI requirements regarding the inspection of longitudinal welds in Class 1 and 2 piping.
TVA proposes to use the alternative examination requirements provided by ASME Code Case N-524,
" Alternative Examination Requirements for Longitudinal Welds in Class 1 and Class 2 Piping Section XI, Division 1."
Code Case N-524 allows examination of longitudinal piping welds at intersecting circumferential welds within the examination boundary of the circumferential weld, rather than examination of one pipe diameter or up to twelve inches of each longitudinal piping weld that intersects a circumferential weld.
Request for relief 3-ISI-3 is consistent with one previously accepted by NRC letter dated September 28, 1994, for use by BFN Unit 3 during its First Ten-Year Inservice Inspection Interval.
Unit:
Second ASME Section XI ISI Inspection Interval, (Start Date:
November 19, 1996)
Systems:
Various ASME Section XI Systems J
Components:
Longitudinal Welds in Class 1 and 2 Piping Code Class:
Class 1 and 2 Examination Category:
B-J, C-F-1, and C-F-2 Item Number:
B9.12, B9.22, C5.12, C5.22, C5.42, C5.52, C5.62, and C5.82 l
Code Requirement:
1989 Edition of ASME Section XI (no l
addenda), Table IWB-2500-1 requires a volumetric and/or surface examination of at least one pipe diameter length, but no more than 12 inches of each longitudinal weld that intersects a circumferential weld required to be examined in j
accordance with Examination Category B-J.
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d REQUEST FOR RELIEF 3-ISI-3 (cont.)
The 1989 Edition of ASME Section XI (no addenda), Table IWC-2500-1 requires a volumetric and/or surface examination to be performed of at least 2.St, where "t" is the thickness of each longitudinal weld that intersects a circumferential weld required to be examined under Examination Categories C-F-1 and C-F-2.
Code Requirement From Which Relief is Requested:
In accordance with 10 CFR 50.55a (a) (3) (i), relief is requested to use ASME Code Case N-524, " Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Section XI, Division 1," in lieu of the 1989 ASME Section XI requirements listed above.
Basis For Relief:
The alternative examination requirements of Code Case N-524 include examination of the subject longitudinal piping welds at intersecting circumferential welds within the examination boundary of the circumferential weld.
The following items summarize the basis for the revised longitudinal piping weld examination boundary:
Longitudinal piping welds are fabricated during the manufacturing process under controlled shop conditions, which produce higher quality welds and more uniform residual stress patterns.
Longitudinal piping welds undergo heat treatment during the manufacturing process which enhances the material properties of the weld and reduces the residual stress created by welding.
l Results of previous weld inspections throughout the e
industry indicate that longitudinal welds have not been a safety concern, and there has been no evidence of longitudinal weld defects compromising safety at nuclear power plants.
Longitudinal welds have not been shown to be susceptible to any particular degradation mechanism.
The areas of a longitudinal weld that would be the most susceptible to a potential failure mechanism would be at the intersection of a field fabricated circumferential weld.
l
- f REQUEST FOR RELIEF 3-ISI-3 (cont.)
Locating longitudinal piping welds can require acid e
etching, eddy current examination, or a combination of methods.
This increases radiological exposure, radwaste generation, and overall cost for performance of ASME Section XI examinations.
4 Alternative Requirements:
As an alternative to the requirements stated in the 1989 Edition (no addenda) of ASMF Section XI, Browns Ferry Nuclear Plant will adopt the provisions of ASME Code Case N-524 for the examination of i
Class 1 and 2 longitudinal piping welds.
l Justification For Granting Relief:
TVA's proposed alternative, Code Case N-524, limits the volumetric and surface examination requirements of the longitudinal weld to the volume or area contained within the examination requirements of the intersecting circumferential weld.
If any degradation associated with a longitudinal weld s
were to occur, it is expected that it would be located at the intersection with.a circumferential weld.
These intersections would be inspected in accordance with the provisions of Code Case N-524.
Code Case N-524 was developed by the ASME Code Committee as an alternative to the requirements specified in ASME Section XI for the examination of longitudinal piping welds.
j The alternate examinations specified in the Code case maintain an acceptable level of quality and safety'and are sufficient to verify the continued structural integrity of longitudinal piping welds.
Implementation Schedule:
Code Case N-524 will be implemented during the Second Ten-Year ISI Inspection Interval for Browns Ferry Unit 3.
Attachment:
ASME Code Case N-524, " Alternative Examination Requirements For Longitudinal Welds in Class 1 and 2 Piping,Section XI Division 1" f
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l CASE N-524 4
cAaEE OF AmsE 80523 MS Paseguas ytasEI. Coos as
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- i see manedsat hess 4r morsnea mw any =semesen meses.
t Case N 834 Ame<3eshe Emmadenden Esquiressames Ier j
!.masiendimal Weide la Class 1 and 2 Piples l
j Sessism II, Divialen 1 l
lagety What alternada requirements may be ap.
i plied to the surface and voimmetric==i=ah of i
lamenedinal piping welds speciSed in Tabis IWB-t 25001,18==>am*H Category B J. Table IWC2500 1, 3'===i==*4a= Categones CF 1 and CF 2 (Emme-ination Catagny CF prior to Winter 19g3 Adden-1 da), and Table IWC-2520, F==iamaan Category C
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G (1974 Edition. Sum-ser 1975 Addenda)?
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Agdy: It is the opeiam of the en==inae that the l
follomas shan apply:
(a) When only a surfeos
.... /. is required, i
=va*H of longimdinal poing welds is not re-l quired beyond those portions of the weids ethin the j
a==i==a = bomadaries af imeammains cerammeer-andal welds, t
(b) When both sminos and wohamstric esandma-tiens are required, -=a==*aa= oflonginediaal piping j
weids is man required beyond thans perdams of ths l
weeds wuhim ths===i==*4a= bonaderies of imeer-sesdag cirammierential weids provided the dallowtag i
requinmems== men.
j (1) Where longinsdinal weeds are s locations are known, =e.n, ' pedded and i
. shall j
be est for both trammeros and perauel Seus at the 1
insorseedse of the weeds and int thes nemeth of los-gitadl mal wekt within the cdreemdseendet WS en-aminados vehemst 1
(2) Where longiendinal weeds are speedied but i
locadoms are unknown, or the suissemos of longhm-disal welds is uncertain, ths===i==*ia= require.
meses shot be seat for both treasserse and pernaal Sees wkhim the endre -e-i===% vehens of laser-l sedas carandmedai weidi.
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TENNESSEE VALLEY AUTHORITY BROWNS' FERRY NUCLEAR PLANT UNIT 3 ASME SECTION XI, INSERVICE INSPECTION PROGRAM, SECOND TEN-YEAR INSPECTION INTERVAL REQUEST FOR RELIEF 3-ISI-6, REVISION 1 Executive Summary:
Pursuant to 10 CFR 50.55a (a) (3) (i), TVA is requesting relief from the ASME Section XI requirements regarding the preparation and submittal of NIS-1 and NIS-2 Summary Reports for inservice inspection and repair and replacement activities for each operating cycle.
TVA proposes to use the alternative reporting documentation described in ASME Code Case N-532 which makes similar information available for review at the plant site and requires submission of an Owners Activity Report at the end of the ISI inspection period (approximately every 3.3 years).
The current requirement is for submission of the NIS-1 and NIS-2 Summary Reports to the NRC within 90 days following each refueling outage Unit:
Second ASME Section XI ISI Inspection Interval, (Start Date:
November 19, 1996)
Systems:
Various ASME Section XI Systems Components:
NIS-l'and NIS-2 Summary Reports Code Class:
1, 2,
and 3 Examination Category:
N/A Item Number:
N/A Code Requirement:
1989 Edition, no addenda, of ASME Section XI, IWA-4800, IWA-6210 ( c), IWA-622 0 (c), (d), and (e), IWA-6230, and IWA-7520 (a) (8) requirements for NIS-1 and NIS-2 summary report preparation and submittal.
(The references to IWA-4000, -6000, and -7000 are the equivalent references from the 1989 Edition of ASME Section XI to the references in Code Case N-532 from the 1992 Edition of ASME Section XI]
Code Requirement From Which Relief _i_s Requested:
In accordance with 10 CFR 50.55a (a) (3) (1), relief is requested to use Code Case N-532, " Alternative Requirements to Repair
. - - _ ~ - ~...
REQUEST FOR RELIEF 3-ISI-6 (cont.)
and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by i
IWA-4000 and IWA-6000,Section XI, Division 1,"
in lieu of the 1989 ASME Section XI requirements listed above.
Basis For Relief:
The 1989 Edition of ASME Section XI requires the preparation and submittal of NIS-1 and NIS-2 Summary Reports within 90 days following the completion of each refueling outage.
This requires resources to be diverted from restart of the applicable unit to preparation of the NIS-1 and NIS-2 Summary Reports within the prescribed time limit.
It also presents an unnecessary time constraint on the licensee for submittal of the Summary Reports to NRC.
Code Case N-532 allows an alternate method for the certification of repairs and replacements in conjunction with the preparation and submittal of an Owner's Activity Report (OAR).
Code Case N-532 requires submittal of the OAR to NRC at the end of each inspection period, which allows flexibility of resources at the end of a refueling outage.
The information required for the OAR-1, " Owner's Activity Report" and Form NIS-2A, " Repair / Replacement Certification Record," provide the same level of assurance and third party certification as the corresponding," NIS-1, " Owner's Report for Inservice Inspections" and NIS-2, Owner's Report for Repairs and Replacements from the 1989 Edition of ASME Section XI.
Code Case N-532 requires that additional information be provided at the end of each inspection period regarding the percentage of examinations completed, which assures both the Owner and NRC that ASME Section XI Code compliance has been achieved.
TVA has concluded that Code Case N-532 provides an equal or superior degree of information related to repairs / replacements and ISI examinations completed when compared to the NIS-1 and NIS-2 required by the 1989 Edition of ASME Section XI.
Alternate Examinations:
For BFN Unit 3 TVA will invoke the requirements of ASME Code Case N-532, Alternative Requirements To Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000;Section XI,
- Division 1," as an alternative to the ASME Section XI requirements listed above.
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REQUEST FOR RELIEF 3-ISI-6 (cont.)
Justification For The Granting Of Relief:
Code Case N-532 was developed by the ASME Code Committee as an alternative to the provisions in ASME Section XI regarding repair /
replacement documentation and inservice inspection summary report preparation.
The requirements currently prescribed by the 1989 Edition, no addenda, of ASME Section XI result in preparation of the NIS-1 and NIS-2 Summary Reports for submittal to NRC within 90 days following each refueling outage.
Code Case N-532 makes similar information available for review at the plant site and requires submittal of an Owner's Activity Report at the end of the ISI inspection i
period.
The information presented in the OAR is more concise than that currently required by the NIS-1 and NIS-2.
The use of Code Case N-532 will not affect the margin of safety achieved through implementation of ASME Section XI for verification of system / structural integrity.
Implementation Schedule:
Code Case N-532 will be implemented during the Second Ten-Year ISI Inspection Interval for Browns Ferry Unit 3.
Attachment:
ASME Code Case N-532, " Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000,Section XI, Division 1"
CASE N-532 CA5ES OF ASMF. SOILER AND PRE 55LJRE VF.SSEL Coot Appeevel Dese: Decemhee 12.1994 See Nument indu for uperatoon and any rumemanen detu.
"ese N 532 (e) The completed Form NIS 2A shall be main.
Alternative Requirements to Repair and tained by the Owner.
Repiscement Documentation Requirements and (f) The Owner shall maintain an inder of Repair /
Inservice Summary Remrt Preparation and Replacement Plans in accordance with IWA-6340.
Submission as Requireu by (WA 4000 and (WA.
The inder shall identify the identificat:on number 6000$
required by (b) above and the inspection interval and Section XI. Division 1 period during which each repair or replacement was completed.
Inquuy: What alternatives may be used to the re-2.0 OWNER'S ACTIVI1Y REPORT quirements of IWA 4910(d) and IWA 6210(e) for PREPARATION AND SUBMITTAL -
completion of Form NIS 2 following repair or re.
placement, and IWA 6210(c) and (d), IWA-6220.
An OWNER'S ACTIVITY REPORT FORM IWA-6230(b), (c), and (d), and IWA.6240(b) for OAR 1 shall be prepared and certified upon com-preparation and submittal of the inservice summary pletion of each refueling outage. Each Form OAR.
report and Form NIS.1?
! prepared during an inspection period shall be sub.
mitted following the end of the inspection period.
Reply: It is the opinion of the Committee that as Each Form OAR 1 shall contain the following:
an alternative to the requirements of IWA 4910(d),
(a) Abstract of applicable examinations and tests IWA 6210(c),(d), and (c),IWA 6220,IWA 6230(b),
with the information and format of Table 1.
(c), and (d), and IWA 6240(b), the following provi-(b) A listing of item (s) with Daws or relevam con-sions may be used. This Case shall be utilized at least ditions that required evaluation to determine ac.
until tl'e end of the inspection period in which it was ceptability for contmued service, whether or not the mvoked.
flaw or relevant condiion was discovered dunng a scheduled exammation or test. The listing shall pro.
1.0 CERTIF1 CATION OF THE REPAIR OR vide the information in the fo-mat of Table 2.
REPIACEMENT fc) Abstract for repairs, replacements and correc.
. (a) The Owner's Repair / Replacement Program tive measures performed which were required due shall identify use of this Case.
to an item containing a flaw or relevant condition (b) A Repair / Replacement Plan shall be prepared that exceeded IWB 3000. IWC 3000. IWD 3000, in accordance with IWA 4140', tad shall be given a
!WE 3000, IWF 3000. or IWL 3000 acceptance en-teria: even though the discovery of the flaw or rete.
unique identification number.
(c) Upon completion of all required activities as.
vant condition that necessitated the repair, replace.
sociated with the Repair / Replacement Plan, the ment or corrective measure, may not have resulted Owner shall prepare a REPAIR / REPLACEMENT from an examination or test equired by this Division.
CERTIFICATION RECORD. FORM NIS 2A.
If acceptance criteria for a particular item is not (d) Form NIS 2A shall be presented to the
$Pecified in this Division, the provisions of IWA.
Inspector for certification.
3100(b) shall be used to determine which repairs, replacements, and corrective measures are required to be included in the abstract. The abstract shall pro.
'All reftfences io lWA 4000 and lWA 6000 uned in ehts Case refer to the 1992 Edsuon.
vide the information in the format of Table 3.
1061
I CASE (cCntinued) l N-532 i
l CASES OF A5MI toller AND PRL55(1tE ST.55EL CODE i
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CASE (continued)
N-532 CA5L5 OF ASME BOTLER A.VD PRES $LRE \\ ESSEL CODE E
FORM OAR 1 OWNER'S ACTIVITY A*.PCRT l
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CASE (csntinued)
N-532 C.* *LS OF ASME 1601LER AND FRESSL:RE VLSSEL CODE
\\
TABLE 1 ABSTRACT OF EXAMINATIONS AND TESTS Total l
Totai Total "stal Esamenations Esamanauens E saminates E samu.auons Creet'ad (%) To Esarnination Aeq,.td for Credited for Credited (%)
Data for ne l
C4W9ery The Interval hisPeno4 for The Pened Interval Remarts TABLE 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Fraw or Relevant Constuos Found j E s aminateci Itavn Item Charactentation Dunn, Scheesied section xl j
Category l
Nummer Oextiotica (IWA 3300) l Esaminauen or Test (Yes or No)
TABLE 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE i
rta.or neievant l
l Condeuen Found i
i
- teoast, Ouring Schedu6ed Reolacemeat, Secte x1 Reoanet Code or Correctree item Ceutiotion E samination or cate s 4acement ee Clau Measure Deunopen of W ort Test (YeVNo)
Como+ete Plan we l
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106a
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ATTACHMENT K
- 1. Relief Request 3-ISI-2, Revision 1, Examination and Testing of Component / Piping Snubbers
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I TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 3 ASME SECTION XI, INSERVICE INSPECTION PROGRAM, l
SECOND TEN-YEAR INSPECTION INTERVAL L
I REQUEST FOR RELIEF 3-ISI-2, REVISION 1 Executive Summary:
Pursuant to 10 CFR 50.55a (a) (3) (i), TVA is requesting relief from the identified ASME Section XI requirements related to examination and testing of snubbers.
TVA proposes to use the examination and testing plans currently defined in the BFN Technical Requirements Manual (TRM) 3.7.4.
The current TRM requirements are consistent with the guidance given in Generic Letter 90-09, " Alternate Requirements For Snubber Visual Inspection Intervals and Corrective Actions," which has been promulgated and approved by NRC.
The ASME Section XI criteria imposes overlapping requirements which do not enhance the quality or safety of
)
the subject snubber examination and testing.
Unit:
Second ASME Section XI ISI Inspection Interval, (Start Date:
November 19, 1996)
Systems:
Various ASME Section XI Systems I
Components:
Component / Piping Snubbers Code Class:
1, 2,
and 3 Examination Category:
N/A Item Number:
N/A Code Requirement:
1989 Edition of ASME Section XI (no addenda) IWF-5300(a) and (b) inservice examination and testing in accordance with the first Addenda to ASME/ ANSI OM-1987, Part 4.
IWF-5400 Repairs and Replacements of snubbers shall be in accordance with the first Addenda to ASME/ ANSI OM-1987, l
Part 4.
i l
IWA-6230 requires inservice inspection summary reports for l
snubbers be filed with the regulatory authority, i
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REQUEST FOR RELIEF 3-ISI-2 (cont.)
IWA-2110 requires Authorized Nuclear Inservice Inspector (ANII) involvement for snubber examination and testing Code Requirement From Which Relief is Requested:
In accordance with 10 CFR 50.55a (a) (3) (1), relief is requested from the ASME Section XI, 1989 Edition (no addenda),
requirement for inservice examinations and tests for snubbers, and repair / replacement examinations and tests of snubbers:
a) IWF-5300 (a) and (b) inservice examination and testing, and implied OM-1987, Part 4, Sections 2.3, Inservice Examination, 2.4, Examination Documentation, and 3.2, Inservice Operability Testing, and 3.3, Testing Documentation, b) IWF-5400 Repairs and Replacements of snubbers shall be in accordance with the first Addenda to ASME/ ANSI OM-1987, Part 4, Sections 1.5.6, Snubber Maintenance or Repair, and 1.5.7, Snubber Modification and Replacement c) IWA-6230, summary reports (for snubbers) d) IWA-2110 (a) (5) and (c), Duties of the Inspector (for involvement for snubber examination and testing)
Basis For Relief:
ASME Section XI Class 1, 2 and 3 equivalent snubbers are examined and tested in accordance with Browns Ferry Nuclear (BFN) Plant Technical Requirements Manual (TRM), TR 3.7.4.
BFN TR 3.7.4 is prepared in accordance with the guidance given by NRC in Generic Letter 90-09.
The scope for snubbers examined and tested in accordance with TR 3.7.4 is not limited by line size'or other applicable code exemptions and includes a numerically greater population of snubbers than the Section XI program.
Examination and testing of the snubbers in accordance with both ASME Section XI and the plant TRM would result in a duplication of effort utilizing different standards and require the preparation of a separate program and associated procedures.
This would result in additional cost and unnecessary radiological exposure.
In addition, the personnel performing snubber visual examinations would also be required to be certified in accordance with the American i
Society of Nondestructive Testing (ASNT) SNT-TC-1A i
" Personnel Qualification and Certification in Nondestructive l
Testing," which is an additional certification as compared i
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REQUEST FOR RELIEF 3-ISI-2 (cont.)'
to the task training qualification required to perform the TRM 3.7.4 required examinations and testing of snubbers.
The existing TRM program for examination and testing of l
snubbers'was promulgated and accepted by NRC.
The implementation of OM-1987, Part 4 would require BFN to l
initiate a snubber examination and testing program that is l
more complicated and expensive to perform, without a j
compensating increase in the level of quality and safety.
Alternate Examinations:
The BFN TRM.3.7.4 requirements will be utilized for the examination and testing of snubbers for preservice, inservice, and repair / replacement activities.
The procedures utilized for these examinations are:
3-SI-4.6.H-1,. " Visual Examination of Hydraulic and Mechanical Snubbers;" 3-SI-4.6.H-2A, " Functional Testing of Mechanical Snubbers;" 3-SI-4.6.H-2B, " Functional Testing of l
Bergen-Patterson Hydraulic Snubbers;" 3-SI-4.6.H-2C,
" Functional Testing of Bergen-Patterson. Torus Dynamic Restraints;" MPI-0-000-SNB002, " Hydraulic Shock and Sway Arrestor Bergen-Patterson Unit Disassembly and Reassembly;"
l and MPI-0-000-SNB004, " Instructions for Removirg and
-Reinstalling Pacific Scientific Mechanical, Bergen-Patterson Grinnell Hydraulic, and Torus Dynamic Restraints."
This i
will include the pin-to-pin area inclusive of applicable snubbers.
Testing of repaired and replaced snubbers to establish a new preservice examination will also be performed in accordance with TR 3.7.4.
Visual examination of repaired and replaced snubbers will be performed in accordance with MPI-0-000-SNB-004, " Instruction for Removing and Reinstalling Pacific Scientific Mechanical, Bergen-Patterson Grinnell Hydraulic, and Torus Dynamic Restraints."
Snubber examination and testing data will be maintained in j
accordance with the requirements of TR 3.7.4, the site l
corrective action program, SSP-3.1, and the implementing procedures (3-SI-4.6.H-1, 3-SI-4.6.H-2A, 3-SI-4.6.H-2B,
)
3-SI-4.6.H-2C, MPI-0-000-SNB002, and MPI-0-000-SNB004).
I' The areas, inclusive of the pins back to building structure and to the component / piping being supported, will remain in the ASME Section XI examination boundary.
i l-
REQUEST FOR RELIEF L-ISI-2 (cont.)
l Justification For The Granting Of Relief:
The current program, as defined by TR 3.7.4, provides for a level of quality and safety equal to or greater than that provided by OM Code 1987, Part 4, and utilizes NRC guidance not incorporated into the OM Code referenced by the 1989 Edition of ASME Section XI.
Examination, testing, repair and replacement of snubbers is currently performed in accordance with TR 3.7.4, which utilizes the guidance provided by NRC in Generic Letter 90-09.
The OM Code referenced by ASME Section XI has a different basis for examination (i.e.,
failure mode groups) and testing plans (10, 37, or 55 percent).
It is impractical to implement both plans because of the resulting duplication of examination and testing efforts and different requirements for snubber quantities subject to examination or test, actually examined and/or tested, and sample expansion requirements.
This would result in additional cost and unnecessary radiological exposure.
The existing i
TRM program (based on GL 90-09) for examination and testing of snubbers has been promulgated and accepted by NRC.
The differences in the two programs could create confusion when selecting test samples, applying acceptance criteria, corrective actions, and examination schedules for failed snubbers.
This situation could increase the possibility of applying the wrong action, thus creating a nonconformance, an inoperability, or a violaticn of TRM requirements.
To eliminate any misinterpretation or confusion in administering overlapping requirements for snubbers, and to remove the possibility of applying contradicting requirements to the same snubber (s), BFN proposes to examine and test snubbers in accordance with BFN TR 3.7.4.
Subarticle IWF-5400 prov.idas the requirements for repair and replacement of snubbers to be in accordance with OM-1987, Part 4.
OM-1987, Part 4, Sections 1.5.6,
" Snubber Maintenance or Repair" and 1.5.7,
" Snubber Modification and Replacement" require repaired and replaced snubbers to meet the visual examination requirements of Paragraph 2.3.1.2 and the operability test requirements of Paragraph 3.2.11.
Section 1.5.6 also requires an evaluation of the maintenance or repair activity and Section 1.5.7 requires a suitability evaluation of the replacement / modified snubber.
TR 3.7.4 (TSR 3.7. 4. 6) requires replacement snubbers and snubbers
l REQUEST FOR RELIEF 3-ISI-2 (cont.)
l l
which have repairs which might affect the functional test results to be tested to meet the functional test criteria prior to installation.
Maintenance procedure MPI-0-000-SNB004 provides visual examination criteria for installation of a snubber after repair or replacement.
The ASME Section XI repair / replacement program at BFN documents the suitability of repairs, IWA-4130 (a) (4 ), and replacements, IWA-7220.
ASME Section XI VT-3 certification required by personnel performing snubber visual examinations is an additional certification as compared with the TRM program training qualifications.
Personnel performing the TRM required visual examinations are " process qualified" to perform the examinations and testing required by the TRM and implemented by the referenced procedures.
This training currently includes a visual test associated with face mask fit and specific training on the acceptance criteria associated with procedure MPI-0-000-SNB004.
Additional " visual acuity" verification for personnel performing snubber visual examinations will include visual acuity requirements that meet ASME Section XI.
The training and documentation of personnel to the visual acceptance criteria, specified in the TRM implementing procedures, provides an acceptable level of quality and safety.
Because relief is sought from the ASME Section XI snubber examination and test requirements, there will be no ASME Section XI snubber examination and test activities to require ANII involvement.
The BFN TRM snubber program does not require the use of an ANII for examination and test requirements.
The ANII will not be involved in the TRM required visual examination or testing activities performed in lieu of the ASME Code requirements.
A snubber program manager provides oversight of the TRM snubber program implementation for both visual examination and functional testing.
This oversight includes both review and evaluation of visual examination and functional testing data to ensure TRM requirements are met.
The snubber program manager provides an acceptable level of quality and safety without ANII involvement in those activities.
ANII involvement in l
other inservice repair and replacement snubber activities, l
as required by IWA-2110(g) and (h) and implemented by BFN's ASME Section XI repair and replacement program, will be maintained.
l
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REQUEST FOR RELIEF 3-ISI-2 (cont.)
l Subarticle IWA-6230 and OM-1987, Part 4, Sections 2.3 and 3.3 provide requirements for ASME Section XI inservice examination and test documentation for snubbers and a summary report of examinations and testing.
Under the l
alternate requirements for snubbers, there will be no ASME Section XI inservice examination and testing to document in a summary report.
TR 3.7.4 is implemented by surveillance instructions 3-SI-4.6.H-1, 3-SI-4.6.H-2A, 3-SI-4.6.H-2B, and 3-SI-4.6.H-2C and maintenance instruction MPI-0-000-SNB004.
These instructions are written and approved in accordance with the TVA Nuclear Quality Assurance Program, include data sheets for documenting the visual examination and functional test data and results, and provide for documentation of nonconforming results and evaluation of those results.
The completed data sheets are QA records and are controlled and maintained in accordance with the BFN QA records program.
These records are available onsite for review and inspection.
The QA records documenting snubber visual examinations and functional tests provide an acceptable level of quality and safety when compared to the requirements of ASME Section XI, and OM-1987, Part 4.
Based on the justification provided above, BFN's alternative program for examination and testing of snubbers in accordance with TR 3.7.4 will provide an acceptable level of quality and safety.
Therefore, pursuant to 10 CFR 50.55a (a) (3) (1), TVA requests that relief be granted from the 1989 Edition of ASME Section XI Code requirements related to inservice examination and testing for snubbers.
Implementation Schedule:
TR 3.7.4 will be implemented during the Second Ten-Year ASME Section XI Inservice Inspection Interval for snubber examination and testing in lieu of the'ASME Code requirements listed above.
Attachment:
Browns Ferry Nuclear Plant, Unit 3 Technical Requirements Manual, Section TR 3.7.4,
" Snubbers," and the associated bases.
I l
\\
i
TR 3.7 PLANT SYSTEMS l
TR 3.7.4 Snubbers LCO 3.7.4 During all MODES of operation, all snubbers shall be OPERABLE.
All safety-related snubbers are listed in plant procedures.
APPLICABILITY:
MODES 1, 2, 3, 4, 5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more NOTE snubber (s) inoperable An engineering evaluation of the on a system that is components which are required to be restrained by the inoperable OPERABLE in the snubber (s) shall be completed if current plant this Condition is entered.
Condition.
A.1 Replace or restore the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable snubber (s) to OPERABLE status and perform an engineering evaluation on the attached component.
9.8 A.2 Declare the attached 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> system inoperable.
(Refer to applicable TS and TRM LCOs).
i l
BFN-UNIT 3 3.7-7 TRM Revision 0
Each safety-related snubber shall be demonstrated OPERABLE by peiformance of the following augmented inservice inspection program and the requirements of this Technical Requirement. These snubbers are listed in plant procedures.
2.
As used in this Technical Requirement, " type of snubber" shall mean snubbers of the same design and manufacturer, irrespective of capacity.
i TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
TSR 3.7.4.1 NOTE Snubbers are categorized as inaccessible or accessible during reactor operations. Each of these categories (inaccessible and i
accessible) may be inspected independently accoroing to the schedule determined by Table 3.7.4-1. The visual inspection interval i
for each type of snubber shall be determined based upon the criteria provided by Table 3.7.4-1. The first inspection interval determined using Table 3.7.4-1 criteria shall be based on the previous inspection interval established by the requirements in effect before Technical Specification Amendment 183 was issued.
Perform visual inspection of required in accordance i
snubber based on the criteria for each with Table category in Table 3.7.4-1 to verify:
3.7.4-1 a.
No visible indications of damage or impaired OPERABILITY; b.
Attachments to the foundation or supporting structure are functional; and c.
Fasteners for the attachment of the snubber to the component and to the l
snubber anchorage are functional.
(continued)
BFN-UNIT 3 3.7-8 TRM Revision 0
i Snubbers TR 3.7.4 TECHNICAL SURVEILLANCE REQUIREMENTS j
SURVEILLANCE FREQUENCY i
j TSR 3.7.4.1 Additionally, snubbers attached to sections (continued) of safety-related systems that have experienced unexpected potentially j
damaging transients since the last inspection l
period shall be evaluated for the possibility l
of concealed damage and functionally tested, if applicable, to confirm OPERABILITY.
Snubbers which appear inoperable as a result of visual inspection shall be classified j
unacceptable, j
A review and evaluation shall be performed 1
and documented to justify continued operation with an unacceptable snubber. If 1
continued operation cannot be justified, the j
snubber shall be declared inoperable.
J TSR 3.7.4.2 Perform an in-place or bench functional test During each
)
of a representative sample of 10% of the REFUELING total of each type of safety-related snubber.
OUTAGE a.
The representative sample selected for functional testing shall include the j
various configurations, operating i
environments, and the range of size and
}
capacity of snubbers within the types; J
i b.
The representative sample should be I
weighed to include more snubbers from
)
severe service areas such as near heavy equipment; C
c.
The stroke setting and the security of fasteners for attachment of the snubbers to the component and to the snubber anchorage shall be verified.
(continued)
BFN-UNIT 3 3.7-9 TRM Revision 0
d Snubbers TR 3.7 4 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY f
TSR 3.7.4.2 The snubber functional test shall verify that:
(continued) a.
Activation (restraining action)is achieved in both tension and l
compression within the specdied range, except that inertia dependent, j
acceleration limiting mechanical snubbers may be tested to verify only that activation takes place in both directions of travel.
b.
Snubber bleed, or release where required, is present in both compression and tension within the specified range.
c.
For mechanical snubbers, the force I
required to initiate or maintain motion of the snubber is not great enough to overstress the attached piping or component during therinal movement, or 1
to indicate impending failure of the snubber.
j d.
For snubbers specifically required not to
]
displace under continuous load, the
{
ability of the snubber to withstand load without displacement shall be verified.
j e.
Testing methods may be used to measure parameters indirectly or i
parameters other than those specified if those results can be correlated to the specified parameters through established methods.
An engineering evaluation shall be made of each failure to meet the test acceptance criteria to determine the cause of the failure.
The result of this analysis shall be used, if applicable, in selecting snubbers to be tested in the subsequent lot in an effort to determine the OPERABILITY of other (continued)
BFN-UNIT 3 3.7-10 TRM Revision 0
. - ~ - -
Snubbers TR 3.7.4 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4
TSR 3.7.4.2 snubbers which may be subject to the same (continued) failure mode. Selection of snubbers for future testing may also be based on the failure analysis.
The discovery of loose or missing attachment fasteners will be evaluated to
- l determine whether the cause may be i
localized or generic. The result of the j
evaluation will be used to select other i
suspect snubbers for verifying the attachment fasteners, as applicable.
i i
1 TSR 3.7.4.3 For each failed snubber, perform in-place or Once for each bench functional test on an additional lot discovery of j
j equal to 10% of the remainder of that type of snubber failure snubber. Testing shall continue until no to meet additional inoperable snubbers are found functional test 3
within subsequent lots or all snubbers of the acceptance original test type are tested or all suspect criteria snubbers identified by the failure analysis have been tested, as applicable. The functional test criteria shall be as specified in TSR 3.7.4.2.
TSR 3.7.4.4 NOTE This testing is independent of the requirements of TSR 3.7.4.3.
i i
For any snubber which fails to lockup or fails Once for each i
to move (i.e., frozen in place), evaluate the discovery of t
i cause.. If caused by manufacturer or design snubber failure deficie'ncy, perform in-place or bench to lockup or i
functional test of all snubbers of the same failure to move design subject to the same defect. The 2
functional test acceptance criteria shall be as specified in TSR 3.7.4.2.
(continued)
BFN-UNIT 3 3.7-11 TRM Revision 0
l Snubbers TR 3.7.4 TECHNICAL SURVEILLANCE REQUIREMENTS 1
\\
i SURVEILLANCE FREQUENCY TSR 3.7.4.5 Perform an engineering evaluation on the Once for each components which are restrained by the discovery of an snubber (s) found inoperable.
inoperable snubber TSR 3.7.4.6 Verify replacement snubbers and snubbers Once prior to which have repairs which might affect the installation in functional test results meet the test criteria of the unit for each TSR 3.7.4.2.
replacement snubber and a.
These snubbers shall have met the each snubber acceptance criteria subsequent to their which has most recent service; and repairs which might affect b.
The functional test must have been functional test performed within 12 months before being results installed in the unit.
BFN-UNIT 3 3.7-12 TRM Revision 0 l
Snubbers TR 3.7.4 Table 3.7.4-1 Snubber Visual inspection Interval i
i NUMBER OF UNACCEPTABLE SNUBBERS Population Column A Column B Column C l
or Category Extend Interval Repeat interval Reduce interval-(Notes 1,2 and 7)
(Notes 3,6 and 7)
(Notes 4,6 and 7)
(Notes 5,6 and 7) i 1
0 0
1 80 0
0 2
)
100 0
1 4
)
150 0
3 8
200 2
5 13 1
300 5
12 25 i
400 8
18 36 i
i 500 12 24 48 i
750 20 40 78 j
1000 or more 29 56 109 i
j Note 1:
The next visual inspection interval for a snubber population or category size
- l shall be determined based upon the previous inspection interval and the l -
number of unacceptable snubbers found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible. These categories snay be examined separately or jointly. However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to j
determine the next inspection interval for that category.
4 Note 2:
Interpolation between population or category sizes and the number of unacceptable snubbers is permissible. Use next lower integer for the value of the limit for Columns A, B, or C if that integer includes a fractional value of unacceptable snubbers as determined by interpolation, i
i
)
BFN-UNIT 3 3.7-13 TRM Revision 0 i
_. ~ _ _ _ _ _ _. _.
_._ __ _ _ _. _ _ _ = _ _
Snubbers TR 3.7.4 Table 3.7.4-1 (Continued)
Snubber Visual inspection Interval Note 3:
If the number of unacceptable snubbrs is equal to or less than the number
)
in Column A, the next inspection interval may be twice the previous interval l
but not greater than 48 months.
Note 4:
If the number of unacceptable snubbers is equal to or less than the number in Column B but greater than the number in Column A, the next inspection
]
interval shall be the same as the previous interval.
Note 5:
If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the 1
previous interval. However, if the number of unacceptable snubbers is less i
than the number in Column C, but greater than the number in Column B, the j'
next interval shall be reduced proportionally by interpolation, that is, the i
previous interval shall be reduced by a factor that is one-third of the ratio of j
the difference between the number of unacceptable snubbers found during j
the previous interval and the number in Column B to the difference in the numbers in Columns B and C.
I Note 6:
The provisions of TSR 3.0.1, 3.0.2, and 3.0.3 are applicable for all inspection intervals up to and including 48 months.
Note 7:
Snubbers which appear inoperable as a result of visual inspections shall be classified unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per the criteria of TSR 3.7.4.2.
Snubbers which have been made inoperable as the result of unexpected transients, isolated damage, or other random events, when the provisions of TSR 3.7.4.5 and TSR 3.7.4.6 have been met and any other appropriate corrective action implemented, shall not be counted in determining the next visual inspection interval.
BFN-UNIT 3 3.7-14 TRM Revision 0
Snubbers L
B 3.7.4 l
TR 3.7 PLANT SYSTEMS l
BASES BACKGROUND Snubbers are designed to prevent unrestrained pipe or component motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping or components as a result of a seirmic or other event initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be OPERABLE during MODES 1,2,3,4, and 5 when the component or system is required OPERABLE.
APPLICABLE Once a snubber is found to be inoperable, an engineering SAFETY ANALYSIS - evaluation shall be performed to determine the effects on the supported component and the OPERABILITY of the affected system (s). Additionally, only a limited amount of time is allowed for the attached system to be considered OPERABLE, since adequate protection during seismic or other events initiating dynamic loads mcy not be provided with the snubber inoperable.
LCO 3.7.4 The number of snubbers on each unit and shared systems at BFN are too numerous to list in this Technical Requirement. The surveillance 4.6.H series lists the components that are required.
APPLICABILITY All MODES are applicable for this requirement since the snubbers affect a wide variety of systems, of which some of the systems are required in one or more MODES of operation. All MODES are covered since every MODE will be applicable for one or more of the supported systems.
However, if a component or system affected by the snubber is not iequired to be OPERABLE, then the snubber is not required to be OPERABLE. During the times the snubbers are not required to be OPERABLE for an inoperable system, the inoperable snubber (s) i I
will be tracked to prevent declaring the system OPERABLE with unanalyzed inoperable snubbers.
BFN-UNIT 3 8 3.7-8 TRM Revision 0
~
l Snubbers B 3.7.4 BASES ACTIONS A.1 and A.2 l
l Because the protection is required only during relatively low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to replace or restore the inoperable s<nubber to OPERABLE status and perform an engineering evaluation on the supported component or declare the supported system incperable. The engineering evaluation is performed to determine whether the mode of failure of the snubber has adversely affected any safety-related component or system.
TECHNICAL A note is provided to indicate that in this Technical Requirement, SURVEILLANCE
" type of snubber" shall mean snubbers of the same design and REQUIREMENTS manufacturer, irrespective of capacity.
An additional not is provided to indicate that each safety-related snubber (listed in plant procedures) shall be demonstrated OPERABLE by performance of the augmented inservice inspectio'n program and requirements of this Technical Requirement. The augmented inservice inspection program includes the following.
To verify snubber OPERABILITY functional test shall be performed during the REFUELING OUTAGES, at approximately 18-month intervals.
These tests will include stroking of the snubbers to verify proper movement, activation, and bleed or release. Ten percent represents an adequate sample for such tests. Observed failures l
on these samples will require an engineering analysis and testing of additional units, if the engineering analysis results in the determination that the failure of a snubber to activate or to stroke (i.e., seized components) is the result of manufacture or design deficiency, all snubbers subject to the same defect shall be functionally tested. A thorough visual inspection of the snubber i
threaded attachments to the pipe or components and the anchorage will be made in conjunction with all required functional tests. The stroke setting of the snubbers selected for functional L
testing also will be verified.
i BFN-UNIT 3 B 3.7-9 TRM Revision 0
Snubbers 1
8 3.7.4 BASES TECHNICAL All safety-related snubbers are also visually inspected for overall SURVEILLANCE integrity and OPERABILITY. The inspection will include REQUIREMENTS verification of proper orientation, adequate fluid ievel if app!! cable, (Continued) and proper attachment of the snubber to piping and structures.
The removal of insulation or the verification of torque values for
{
threaded fasteners is not required for visual inspections.
The visual inspection frequency is based upon maintaining a constant level of snubber protection. The number of inoperable snubbers found during a required inspection determines the time interval for the next required inspection. Inspections performed i
before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25 percent) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.
When the cause of the rejection of a snubber in a visual inspection is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible and OPERABILITY verified by inservice functional testing, if applicable, that snubber may be reclassified as OPERABLE. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration, inspection types may be established based on design features, and installed conditions which may be expected to be generic. Each of these inspection types is inspected and tested separately unless an engineering analysis indicates the inspection type is improperly constituted. All suspect snubbers are subject to inspection and testing regardless of inspection type.
i i
Exemption from Visual Inspection or Functional Tests:
Permanent or other exemptions from visual inspections and/or j
functional testing for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and if applicable snubber life destructive testing was performed to qualify 1
snubber OPERABILITY for the applicable design conditions at j
either the completion of their fabrication or at a subsequent date.
Snubbers so exempted shall continue to be listed in the plant instructions with footnotes indicating the extent of the exemptions.
BFN-UNIT 3 8 3.7-10 TRM Revision 0
- ~
j Snubbers B 3.7.4 BASES i
TECHNICAL Snubber Service Life Program:
i SURVEILLANCE REQUIREMENTS The service life of snubbers may be extended based on an l
(continued) evaluation of the records of functional tests, maintenance history, and environmental conditions to which the snubbers have been exposed.
The following will be implemented by the augmented inservice inspection program:
TSR 3.7.4.1 Visual inspections:
Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these categories (inaccessible and accessible) may be inspected independently according to the i
schedule determined by Table 3.7.4-1. The visual inspection Interval for each type of snubber shall be determined based upon the criteria provided in Table 3.7.4-1 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Technical Specification amendment No.183.
Visual Inspection Acceptance Criteria:
Visual inspections shall verify that (1) the snubber has no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are functional, and (3) fasteners for the attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per the criteria of TSR 3.7.4.2.
A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber. If continued operation cannot be justified, the snubber shall be l
declared inoperable.
I BFN-UNIT 3 8 3.7-11 TRM Revision 0 i
Snubbers B 3.7.4 BASES i
TECHNICAL TSR 3.7.4.1 (continued)
SURVEILLANCE l
REQUIREMENTS Additionally, snubbers attached to sections of safety-related systems that have experienced unexpected potentially damaging transients since the last inspection period shall be evaluated for i
the possibility of concealed damage and functionally tested, if applicable, to confirm OPERABILITY. Snubbers which have been 1
made inoperable as the result of unexpected transients, isolated damage, or other random ~ events, when the provisions of TSR 3.7.4.5 and 3.7.4.6 have been met and any other appropriate corrective action implemented, shall not be counted in determining L
the next visual inspection interval.
1 l-TSR 3.7.4.2. TSR 3.7.4.3. and TSR 3.7.4.4 Functional Test Schedule, Lot Size, and Compcsition:
During each A.EFUELING OUTAGE, a representative sample of '
f 10% of the total of each type of safety-related snubbers in use in the plant shall be functionally tested either in place or in a bench test. The representative sample selected for functional testing shall include the various configurations, operating environments, and the range of size and capacity of snubbers within the types.
The representative sample should be weighed to include more snubbers from severe service areas such as near heavy equipment. The stroke setting and the security of fasteners for attachment of the snubbers to the component and to the snubber anchorage shall be verified on snubbers selected for functional tests.
Functional Test Acceptance Criteria:
The snubber functional test shall verify that:
- a. Activation (restraining action) is achieved in both tension and compression within the specified range, except that inertia dependent, acceleration limiting mechanical snubbers may be tested to verify only that activation takes place in both directions of travel.
b.
Snubber bleed, or release where required, is present in both compression and tension within the specified range.
BFN-UNIT 3 B 3.7-12 TRM Revision 0
Snubb:rs B 3.7.4 BASES TECHNICAL TSR 3.7.4.2. TSR 3.7.4.3. and TSR 3.7.4.4 (continued)
SURVElLLANCE REQUIREMENTS c.
For mechanical snubbers, the force required to initiate or maintain motion of the snubber is not great enough to overstress the attached piping or component during thermal movement, or to indicate impending failure of the snubber.
d.
For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.
e.
Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.
Functional Test Failure Analysis and Additional Test Lots:
An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The result of this analysis shall be used, if applicable, in selecting snubbers to be tested in the subsequent lot in an effort to determine the OPERABILITY of other snubbers which may be subject to the same failure mode. Selection of snubbers for future testing may also be based on the failure analysis. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic. The result of the evaluation will be used to select other suspect snubbers for verifying the attachment fasteners, as applicable.
For each snubber that does not meet the functional test acceptance criteria, an additional lot equal to 10 percent of the l
remainder of that type of snubbers shall be functionally tested.
Testing shall continue until no additional inoperable snubbers are found within subsequent lots or all snubbers of the original functional test type have been tested or all suspect snubbers identified by the failure analysis have been tested, as applicable. If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency, all snubbers of the same design subject to the same defect shall be functionally I
tested. This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional test acceptance criteria.
BFN-UNIT 3 B 3.7-13 TRM Revision 0 l
Snubbers B 3.7.4 BASES TECHNICAL TSR $.7.4.5 SURVEILLANCE REQUIREMENTS Functional Test Failure - Attached Component Analysis:
(continued) f Fpr the snubber (s) found inoperable, an engineering evaluation sblall be performed on the components which are restrained by the snubber (s). The purpose of this engineering evaluation shall be to determine if the components restrained by the snubber (s) were adversely affected by the inoperability of the snubber (s), and in
~
order to ensure that the restrained component remains capable of meeting the designed service.
TSR 3.7.4.6 Functional Testing Of Repaired and Spare Snubbers:
Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs ~L,h might affect the functional test results shall meet the functional test criteria before
\\
installation in the unit. These snubbers shall have met the acceptance criteria subsequent to their most recent service, and the functional test must have been performed within 12 months before being installed in the unit.
REFERENCES 1.
BFN Technical Specifications (version prior to standardized version) i BFN-UNIT 3 8 3.7-14 TRM Revision 0 L
I A
b
i 7-ATTACHMENT A l
BFN Unit 3 Inservice Inspection Classification Boundary Diagrams i
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