ML20211C263
| ML20211C263 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 08/13/1999 |
| From: | Brooks M, Hambleton R, Sheffel B DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20211C249 | List: |
| References | |
| PROC-990813-01, NUDOCS 9908250192 | |
| Download: ML20211C263 (157) | |
Text
I ISI-NDE Program Rev.2 Change 0 DETROIT EDISON COMPANY 2000 SECOND AVENUE DETROIT, MICHIGAN 48226 INSERVICE INSPECTION NONDESTRUCTIVE EXAMINATION (ISI NDE) PROGRAM (PLAN)
FOR FERMI 2 POWER PLANT DOCUMENT NO. ISI NDE PROGRAM FERMI 2 NUCLEAR OPERATIONS CENTER 6400 N. DIXIE HIGHWAY NEWPORT, MICHIGAN 48166 DATE OF COMMERCIAL OPERATION 123-88 Second Interval Start February 17,2000 Projected Cosupletion Date February 16,2010 k8!99 Reviewed by:
I M 8!/8/99 Prepared by:
M. A. Brooks
/Date R. M. Hambleton
/Date NDE LevelIII Lead ISI Engineer
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Direc erformance Engineering
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l ARMS -INFORMATION SERVICES l
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Date approved:
Release authorized by-Change numbers incorporated:
l l
DSN ISI-NDE Program Change O
Rev 2
Date DTC TM PLAN File 1715.04 Recipient 9908250192 990819 PDR ADOCK 05000341 p
l ISI NDE Prograra Revision 2 ChangeO Part A Page I l
PART-A INSERVICE INSPECTION-NONDESTRUCTIVE EXAMINATION (ISI-NDE) PROGRAM (PLAN)
FOR PIPING, COMPONENTS, AND INTEGRAL ATTACllMENTS i
i 1
ISI NDE Program Revision 2 Change O Pan A Page 2 1.0 Applicable Code The Inservice Inspection Program (Plan) for Nondestructive Examination (NDE) of Class 1, 2 and 3 components is implemented in accordance with the requirements of Subsections IWA, IWB, IWC, LWD, and IWF Section XI, Division 1 of the ASME Boiler and Pressure Vessel Code,1989 Edition. All references in this document (Part A) to Section XI,- the Code, Categories, item Numbers, etc. - refer to the 1989 Edition with no Addenda unless otherwise noted. Containment inspections per subsection IWE are not included as part of this
{
program.
2.0 Program Description The NDE Program (Plan) developed herein will utilize Inspection Program B, (IWA-2420),
Tables IWB-2412-1, IWC-2412-1, and IWD-2412-1 for the second 120 month inspection interval. Per 10CFR50.55a(g)(5)(iii), where the requirements of the governing Code Edition or Addenda are determined to be impractical, within the limitations of design, geometry and materials of construction of the components, specific relief requests have been written. These j
are referenced in the applicable NDE Program Tables and included in Section 4.0 of this program (Part A). If during the course ofimplementation the need for additional reliefis identified, it will be requested at a later date.
l The Second Ten-year Interval will begin on February 17,2000 and is planned to include six refueling and inspection outages (reference ISI evaluation 98-005).
The extent of examination for ASME Class-2 pipe welds has been modified for the second ten-year interval as described in the 1989 Edition of ASME Section XI, Table IWB-2500-1, Category C-F. The selection ofindividual welds to be examined on each Class 2 system is based on the inspection philosophy described in Fermi UFSAR Section 5.2.8.8 along with the prorating methodology described in the 1989 Code (reference ISI Evaluation 99-053).
l The systems or portions of systems subject to the examination requirements of the ISI-NDE l
Program for Fermi 2 and the associated Class 1,2, and 3 boundaries are documented on ISI l
Classification Boundary Drawings. The ISI Classification Boundary Drawings are listed in Table A-5-5.1.
Section 5.0, Table A-5-5.2 also li-ts Fermi p Inservice Inspection Isometric Drawings which have been developed for all vessels, piping, pumps, and valves which require volumetric and/or surface examination by this program.
Revision 2 Change 0 i
Part A Page 3 The system inservice inspection classifications, developed specifically to define the extent to which ASME Section XI requirements will be applied, differ somewhat from the ASME Section 111 design classifications. Primarily, these differences occur because systems, or portions of systems, were optionally upgraded in design. In addition, ISI classifications are limited to systems important to safety, which contain water, steam, or radioactive materials.
3.0 Exemptions Certain components (or parts of components) and the associated supports are exempt from the examination requirements of ASME Section XI, Table IWB-2500-1, per provisions contained in the Code. This section lists those " Code Allowed Exemptions" as applicable to the Fermi 2 Inservice Inspection Program.
3.1 Class-1 Components exempt from volumetric and surface examination requirements of IWB-2500 of the Code:
Exemption EX-Al-1 ASME Section XI Code Paragraph: IWB-1220(a)
" Components that are connected to the reactor coolant system and part of the reactor coolant pressure boundary and that are of such size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under normal plant operating conditions is within the capacity of makeup systems which are operable from on-site emergency power."
Justification:
The maximum size line break for Fermi that can be made up by the reactor coolant makeup system has been calculated to be 1.44 inches inside diameter for liquid carrying lines and 2.88 inches inside diameter for steam carrying lines.
Exemption EX-Al-2 ASME Section XI Code Paragraph: IWB-1220(b)(1) and (2)
" Piping of 1 inch nominal pipe size and smaller and components and their connections in piping of 1 inch nominal pipe size and smaller."
t
I ISI NDE [Yogram Revision 2 Change 0 Part A Page 4 Exemption EX-Al-3 ASME Section XI Code Paragraph: IWB-1220(c)
" Reactor Vessel head connections and associated piping,2 inches nominal pipe size and smaller, made inaccessible by control rod drive penetrations."
3.2 Class-2 Components exempt from the inservice examination requirements ofIWC-2500 of the Code:
Exemption EX-A2-1 ASME Section XI Code Paragraph: IWC-1221(a),IWC-1222(a)
Vessels, piping, pumps, valves, and other components NPS 4 and smaller.
Exemption EX-A2-2 ASME Section XI Code Paragraph: IWC-1221(c), and IWC-1222(b)
Component connections NPS 4 and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size.
Exemption EX-A2-3 ASME Section XI Code Paragraph: IWC-1221(f), and IWC-1222(d)
Piping and other components of any size beyond the last shutoff valve in open-ended portions of systems that do not contain water during normal plant operating conditions.
Exemption EX-A2-4 ASME Section XI Code Paragraph: IWC-5222(d)
Open ended portions of discharge lines beyond the last shutoff valve in non-closed systems, demonstration of an open flow path test shall be performed in lieu of the system hydrostatic test."
~
ISI NDE Program Revision 2 Change O Part A Page 5 Exemption EX-A2-5.
ASME Section XI Code Paragraph: IWC-1222(d)
Vessels, piping, pumps, valves, other components, and component connections of any size in systems or portions of systems other than RiiR, ECC, and CIIR, that operate (when the system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200"F.
3.3 Class-3 Components exempt from the inservice inspection requirements ofIWD-2500 of the Code:
Exemption EX-A3-1 ASME Section XI Code Paragraph: IWD-1220.1 Integral attachments to supports and restraints to components that are 4 inches nominal pipe size and smaller within the system boundaries of examination categories D-A. D-B, and D-C of Table IWD-2500-1 shall be exempt from the visual examination VT-3.
Exemption EX-A3-2 j
ASME Section XI Code Paragraph: IWD-1220.2(a) and (b)
" Integral attachments of supports and restraints to components exceeding 4 inches nominal pipe size may be exempted from the visual examination VT-3 of Table IWD-2500-1 provided:
(a) The components are located in systems (or portions of system) whose function is not required in support of reactor residual heat removal, containment heat removal, and emergency core cooling; and (b) The components operate at a pressure of 275 psig or less and at a temperature of 200 degrees F or less."
Exemption EX-A3-3 ASME Section XI Code Paragraph: IWD-5223(d)
For open ended portions of discharge lines beyond the last shutoff valve in nonclosed systems, confirmation of adequate flow during system operation shall be acceptable in lieu of system hydrostatic test.
l
151 NDE Program Revision 2 Change 0 Part A Page 6 Exemption EX-A3-4 ASME Section XI Code Paragraph: IWD-5223(e)
Open ended vent and drain lines from components extending beyond the last shutoff valve and open-ended safety of relief va'.ve discharge lines shall be exempt from hydrostatic test.
4.0 Relief Requests Relief Requests are included where specific requirements of ASME Section XI are determined to be impractical. Each Relief Request is written in accordance with the format guidelines provided in Section 4.1. Individual Relief Requests are included in Section 4.2.
As noted in the INTRODUCTION, Section 4.2 is subject to change throughout the inspection interval. If examination requirements in this program plan are determined to be impractical during the course of the interval, additional or modified Relief Requests will be submitted in accordance with 10 CFR 50.55a (g)(5)(iii).
4.1 Relief Request Format Each Relief Request will include the following sections:
4.1.1 COMPONENT DESCRIPTION: The component description will include:
A general description of the component (s) addressed by the Relief Request, o
o The applicable Plant identification System (PIS) number (s) which uniquely identify the plant system and specific component (s) within the system, and o A quantity of components if the Relief Request covers more than ten (10) components.
4.1.2 ASME CODE CLASS: The ISI classification, Class 1,2,3, as identified on the ISI Classification Boundary Drawings will be listed.
l l
4.1.3 ASME SECTION XI REQUIREMENT: The impractical ASME Section XI l
requirement (s) will be listed. To the extent possible, subparagraphs, individual footnotes, or specific Item Numbers will be cited.
4.1.4 HASIS FOR RELIEF: Information to support Detroit Edison's determination that the Code requirement is impractical will be provided. The following data will be provided,if applicable:
i
ISI NDE Program Revision 2 Change 0 Part A Page 7 A physical sketch if the component (s) are not accessible for examination.
Detailed technical information (an engineering justification) supporting proposed alternate scope of examination, examination method, or acceptance standard.
A description of the proposed alternative examination's impact on plant safety and justi6 cation of any change in the overall level of plant safety.
A justi6 cation of any change in the overall level of plant safety ifit is not possible to perform alternative examination (s).
Reference to the regulatory basis paragraph.
4.1.5 ALTERNATE EXAMINATION: Any alternate examination (s) that are proposed will be identi6ed. Both alternate examination (s) that are performed in lieu of the Section XI examination requirement (s) and alternate examination (s) that supplement partially completed Section XI examination requirement (s) will be identined. The description of the alternate examination (s) will include a statement describing the extent and frequency of examination, the acceptance standard, and whether deferral of
- inspection to the end of the interval is requested.
4.1.6 APPLICABLE TIME PERIOD: A statement identifying the time period during the inspection interval for which reliefis requested will be included.
4.1.7 Processing and Implementation of NRC requests for reliefis as follows:
Proposed Alternatives to requirements of ASME Section XI may only be used when authorized by the Director of the OfHee of Nuclear Reactor Regulation. The applicant shall de nonstrate per 10CFR50.55a(a)(3) that:
(i) The proposed alternatives would (ii) Compliance with the specified requirements of provide an acceptable level of this section w ould result in hardship or unusual quality and safety equivalent to the or difTiculty without a compensating increase in the required examination level of quality and safety.
Furthermore, if the licensee determines that conformance with certain code requirements is impractical for is facility, the licensee shall notify the NRC as speci6ed in 10CFR50.55a (a)(g)(5)(iii).
r-ISI NDE Prograra Revision 2 Change 0 Part A Page 8 j
Impractical-would require a change in component design to permit the required examination.
- 1. Prepare the relief request in accordance with Part 4.0 of the ISI-NDE Program in sufficient detail to demonstrate one of the above positions and process in accordance with Fermi 2 procedures.
2.
Initiate a Licensing Change Request (LCR)
- 3. Initiate a Preliminary Evaluation (PE)
Note: A relief request is a change to a procedure or assumption made in the UFSAR because j
the UFSAR states and assumes compliance with Section XI. The relief request is a change j
from that assumption. Detroit Edison does not perform the safety evaluation because a relief request requires NRC approval in accordance with the specific regulatory basis paragraph of 10CFR50.55a.
- 4. The Relief Request must be submitted to, reviewed by, and authorized by the Director of the Office of Nuclear Reactor Regulation based on licensee demonstration of the appropriate regulatory provision noted above, prior to its implementation.
4.
Upon receipt of the NRC acceptance, complete the LCR and incorporate the relief requests into the ISI Program.
Note: Where Code requirements are determined to be impractical because of component design, geometry, or materials of construction, NRC acceptance must be granted within twelve months of the expiration of the interval in which the requirement was determined to be impractical as specified in 10 CFR 50.55a(g)(5)(iii).
4.2 Relief Request (s)
The Relief Requests submitted for NRC r.pproval with the second interval program are included in this section.
RELIEF l
REQUEST NUMBER GENERAL DESCRIPTION RR-Al
~ Class 1 Reactor Pressure Vessel Welds Resubmitted at end of first interval (NRC Letter 99-0038) l N_
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r; ISI NDE Program Revision 2 Change O p
Part A Page 9 RELIEF REQUEST NUMBER GENERAL DESCRIPTION l
RR-A6 RPV Nozzle weld partial exam coverage because of configuration l
Resubmitted at end of first interval (NRC Letter 99-0041) l l
RR-A18 Use of ASME Code Case N-546 (Alternative Requirements for -
Qualification of VT examiners)
RR-A19 Alternative Pressure Test Requirements for HPCI i
RR-A23 Limited Examination of Class 1 Piping Welds RR-A25 Alternative Requirements for Examination of RPV Shell Circumferential Welds (BWRVIP-05), Submitted at the end of first interval (NRC Letter 99-0024)
- RR-A26 Examination of Extension of Containment Welds -(ISI Evaluation 99-014)
RR A27 Examination Method for Category B-G-1 Reactor Pressure Vessel Closure Head Nuts RR-A28 IWA 5250(a)(2) Corrective Measures for Leakage at Bolted Connections RR-A29 Sample Expansion Requirements for Austenitic Stainless Steel Welds
)
RR-A30 Application of Utility PD1 Qualified Ultrasonic Examination Personnel and Procedures f
l Change 0 i
Part A Page 10 l
SECOND INTERVAL RELIEF REQUEST RR-Al COMPONENT FUNCTION / DESCRIPTION:
Pressure Retaining Reactor Pressure Vessel (RPV) Shell Welds (See Table 1)
SYSTEM:
Reactor (B11)
ASME CODE CLASS:
Class 1 ASME SECTION XI REOUIREMENTS:
Subsection IWB, Table IWB 2500-1, Examination Category B-A, item No.'s Bl.10 through Bl.40, require volumetric examination of RPV weld and base material regions described in figures IWB-2500-1 through 2500-3 for pressure retaining welds in the reactor pressure vessel each inspection interval.
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(g)(5)(iii) Detroit Edison is requesting relief from ASME Section XI requirements to examine essentially 100% of accessible Category B-A weld lengths because within the limits of RPV design it is impractical to do so. Detroit Edison believes that the alternatives specified provide an acceptable level of quality and safety.
Relief Request RR-Al documented limitations based on both the installed ultrasonic examination system, which used pole tracks for scanning, and part geometry. During RF02 Fermi implemented the use of an automated examination system that uses a magnetic wheel scanning device which maximizes coverage to the extent possible using current technology.
Limitations to automated scanning of RPV shell welds due to the examination system have been eliminated. Current limitations are based only on RPV configuration or interference from other components as described in the " Alternatives" section below.
Reactor Vessel Ultrasonic Examination techniques meet the requirements of ASME Section XI; ASME Section V, Article 4; and Regulatory Guide 1.150. Detroit Edison believes that the alternative examinations proposed satisfy the intent of the ASME Code within the limits of accessibility for examination inherent to the BWR design. Table 1 identifies the welds with limitations and the cause of the limitation (see also attached figures). The extent of examination is reported in accordance with ASME Section V.
1 ISI NDE Program Revision 2 Change O Part A Page 11
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ALTERNATIVES:
)
Welds 1-319A,1-319C,1-319E, & 1-319G The four listed top head weld exams were examined for most of the weld length during the first interval. They are limited because of a lifling lug positioned on each weld. Because of the physical access limitations it is impractical to examine the full volume of these welds for their entire length.
The Fermi proposed alternative for the ASME Code exam performance is partial examination for these welds. For the weld volume that is partially scanned, the ultrasonic examination covers the most critical area at the inside surface of the head. The areas of highest stress on thd outside surface in the area of the limitation (lifting lug attachment welds) receive a surface examination per Category B-H. The alternative of partial examination combined with the surface exam yields similar results to a full examination.
Because of the substantial coverage obtained by the partial ultrasonic examination and the surface examination of the interfering lug / welds, along with the low empirical probability of reactor vessel weld failure, Detroit Edison considers the proposed alternative examination to provide an acceptable level of quality and safety.
Inaccessible Bottom Head Welds Welds 5-306 and 2-306A through 2-306G The access restrictions caused by the CRD penetrations and RPV support skirt make it impractical to perform a meaningful ultrasonic examination of these welds with current technology. For the inaccessible RPV bottom head welds, the proposed alternatives include a combination ASME Section XI Code required leakage inspections and monitoring of drywell leakage during operation.
Reasonable assurance of structural integrity is maintained because the welds received volumetric and surface NDE to verify that no deleterious material or processing defects were present at the l
time of fabrication. The welds are physically located at the bottom of the reactor vessel, below the withdrawn control rod blades. There is also more than 170 inches of water from the bottom -
of the active fuel height to the weld location. This physical arrangement reduces the neutron fluence and the coincident material degrading impacts significantly, when compared to RPV l
beltline welds that are inspectable. The same CRD penetrations that prevent the examination of the welds would also serve to prevent rapid propagation of a large defect by providing a crack i
arrest point.
Because of the visual inspections (VT-2 ) and leakage monitoring performed, physical access limitations, reasonable assurance of structural integrity for these welds, and the low empirical probability of reactor vessel weld failure, Detroit Edison considers the proposed alternative to provide an acceptable level of quality and safety.
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Revision 2 Change 0 j
Part A Page 12 Weld 13-308 The RPV shell to flange weld exam is limited due to vessel flange configuration, which makes it impractical to examine the full volume of the weld. The Code allowed alternative exam of ASME Section V, Article 4, T441.5.1 (Longitudinal exam from the flange) was performed I
during RF06 but this exam was also limited because of the RPV stud holes. Even with this Code allowed alternative, it is not possible to obtain full volume coverage even when scanning is performed from both sides of the weld for 360 degrees.
The Fenni proposed alternative is a partial exam from the shell side combined with the longitudinal wave exam from the flange surface. As shown in Figure 3, the proposed alternative partial exam performed from the shell side provides significant coverage of the ID surface where flaws would be most likely to originate. A significant portion of full weld volume is also covered by the longitudinal exam from the flange surface. Based on physical limitations, the coverage achievable by the alternative examinations, and the low empirical probability of reactor vessel weld failure, Detroit Edison considers the proposed alternative examination to provide an acceptable level of quality and safety.
APPL lCAHl.E TIME PERIOD:
Reliefis requested fbr the second 10-year inspection interval.
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ISI NDE Program Revision 2 Change O Part A Page 13 RR-Al Table 1 LIMITED EXAMINATIONS Category Weld Percentage Limitation Description
/ Item Identification Complete B-A/B l.21 5-306 Inaccessible Bottom head CRD Penetrations and llead Cire. Weld Skin Attachment Weld (Dollar Plate)
B-A/Bl.22 1319A
-73.6%
Top Head Lifting Lugs lid. Merid. Weld B-A/B l.22 1-319C
~70%
Top ilead Lifting Lugs lid. Merid. Weld B-A/Bl.22 1-319E
~72%
Top 11ead Lifting Lugs lid. Merid. Weld B-A/B l.22 1-319G
~71.3%
Top llead Litling Lugs lid. Merid. Weld B-A/B l.22 2-306A Inaccessible Bottom head CRD Penetrations and lid Merid. Weld Skirt Attachment Weld B-A/B l.22 2-306B Inaccessible Bottom head CRD Penetrations and lid. Merid. Weld Skirt Attachment Weld B-A/B l.22 2-306C Inaccessible Bottom head CRD Penetrations and lid. Merid. Weld Skirt Attachment Weld B-A/B l.22 2-306D Inaccessible Bottom head CRD Penetrations and lid. Merid. Weld Skirt Attachment Weld B-A/B l.22 2-306E Inaccessible Bottom head CRD Penetrations and lid. Merid. Weld Skin Attachment Weld B-A/B l.22 2-306F Inaccessible Bottom head CRD Penetrations and lid. Merid. Weld Skirt Attachment Weld B A/Bl.22 2-306G Inaccessible Bo: tom head CRD Penetrations and lid. Merid. Weld Skin Attachment Weld B-A/B l.30 13-308 54 %
RPV Flange Configuration (coverage Shell to Flange augmented by scan from flange seal surface)
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1 ISI NDE Program Revision 2 ChangeO Part A Page 16 FERMI 2 RPV ULTRASONIC SCAN COVERAGE LIMITATIONS I
RR-Al Figure 3 Figure: Weld 13-308 267 3/8" bolt circle FE 17/8" A
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Code Exam Volume A,B,C,D (Weld + 1/2T) 251"lD Limited due to Flange Radius i
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3 ISI NDE Program Revision 2 Change O Part A Page 17 I
l SECOND INTERVAL RELIEF REQUEST
)
RR-A6
)
COMPONENT FUNCTION / DESCRIPTION:
Pressure Retaining Reactor Pressure Vessel (RPV) Nozzle to Shell Welds (See Table 1)
)
SYSTEM:
Reactor (B11)
ASME CODE CLASS:
Class 1 ASME SECTION XI REOUIREMENTS:
Subsection IWB, Table IWB 2500-1, Examination Category B-D, item Nos. B3.90 and B3.100 require volumetric examination of RPV nozzle-to-shell welds and base material regions as shown in figure 2500-7(b).
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(g)(5)(iii) Detroit Edison is requesting relief from ASME Section XI requirements to examine essentially 100% of accessible Category B-D nozzle welds, because within the limits of design and accessibility it is impractical to do so.
Relief Request RR-A6 only documented ultrasonic examination limitations based on interference caused by proximity to other nozzles. Other limitations have been identified during the performance of examinations during the first interval.
The primary limitation to full ASME Code volumetric coverage is nozzle configuration. The nozzle type used in the Fermi 2 reactor is a flanged nozzle as shown in Figure 1. This type of nozzle provides the best access for inspection of the nozzle types permitted in the ASME Code as shown in the Figures ofIWB-2500-7. The Code required volume (t/2) extends into the nozzle outside blend radius. The curve of the radius section hinders the ability of transducers to maintain contact with the nozzle and also changes the effective beam angle. This limitation results in a typical maximum composite coverage of all Code required scans (0 { Longitudinal},
45, and 60 { Parallel & Transverse} degree) between 60% and 70% depending on nozzle diameter and thickness. The maximum obtainable coverage is achieved by the 60-degree transverse (T) j scan. Essentially all of the weld and heat affected zones are covered by this angle beam scan for the entire weld circumference on most nozzles. Typical scan limitations are shown in Figures 2A through 2C. The estimated volumetric coverage obtained is reported in Table 1.
i Another limitation to full ASME Code volumetric coverage is the vessel taper at the bottom head to lower shell course weld. This geometric condition prevents full coverage of the bottom side of i
ISI NDE Program Revision 2 Change 0 Part A Page 18 the two jet pump instrumentation nozzles and the two recirculation suction nozzles. Composite coverage for these welds remains above 60%. This limitation also impacts the nozzle inner radius coverage for the two core spray nozzles as reported in Table 1.
The limitation originally described in RR-A6 of this relief request indicated a limitation of 46 degrees or 12.8% of the full circumference for 2 of 6 feedwater nozzles based on automated examination equipment accessibility. The examinations were performed manually and the limitation was less than originally described and accepted (see Figure 3). A part of the scan path was able to be performed for the full circumference. Additionally, Fermi examines these feedwater nozzles as specified in NUREG 0619 to detect cracking in the nozzle inner radius and bore areas where cracks have previously been detected in other llWRs. These exams were fully completed and no service related flaws have been detected.
i All nozzle forgings received ultrasonic examination during manufacture and the nozzle to shell welds were subject to radiographic examination during fabrication of the reactor pressure vessel.
All of the nozzle welds requiring volumetric examination by AShfE Section XI have been completed during the first ten-year inspection interval and no service related defects have been detected. The nozzle inner radius ultrasonic examination techniques used at Fermi performed scanning from the blend radius; however, since this technique was designed to detect internal surface defects no credit has been taken for those exams.
Reactor Vessel Ultrasonic Examination techniques meet the requirements of ASME Section XI; ASME Section V, Article 4; and Regulatory Guide 1.150. Detroit Edison believes that the extent of examinations completed satisfy the intent of the ASME Code and 10 CFR 50.55a(g)(4) within the limits of accessibility for examination inherent to BWR pressure vessel design.
ALTERNATIVES:
Perform examination of the ASME Code volume to the extent practical.
APPLICAHLE TIME PERIOD:
Reliefis requested for the second 10-year inspection interval.
l
ISI NDE Program Revision 2 Change 0 Part A Page 19 I
RR-A6 Table 1 LIMITED EXAMINATIONS Category Weld Estimated Limitation Description
/ Item Identification Coverage B-D/B3.90 8-316A, through 69.1 %
Nozzle Blend Radius Nozzle Weld 8-316-D B-D/B3.90 4-316A
-60%
Nozzle Blend Radius and Nozzle Weld 4-316D instrumentation Nozzles B D/B3.90 4-316B, C 64.1 %
Nozzle Blend Radius Nonle Weld E. & F B-D/B3.90 14-316A 68.9 %
Nozzle Blend Radius Nozzle Weld 14-316B B-D/B3.90 15-315 68 %
Nozzle Blend Radius Nonle Weld B-D/B3.90 13-314A 66.7 %
Nozzle Blend Radius
- Nozzle Weld through 13-314K B-D/B3.90 5-314 A 65.6 %
Nozzle Blend Radius and Bottom llead Nonle Weld 5-314B to Shell Taper B-D/B3.90 19-314A 63.1 %
Nozzle Blend Radius and Bottom llead l
Nonle Weld 19-314B to Shell Taper B-D/B3.90 2-318 61.4 %
Nozzle Blend Radius Nozzle Weld B D/B3.90 4 3184 62 %
Nozzle Blend Radius Nozzle Weld 4-318B l
B-D/B3.100 19-314A 30.2%
Bottom llead to Shell Taper l
Nonle Weld 19-314B i
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Part A Page14 FERMI 2 ULTRASONIC SCAN COVERAGE LIMITATIONS l
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ISI NDE Program Revision 2 Change 0 Part A Page 25 SECOND INTERVAL RELIEF REQUEST RR-A18 COMPONENT FUNCTION / DESCRIPTION:
Class 1,2, & 3 Pressure Retaining Piping & Compc"ents SYSTEM:
All systems included in the ISI NDE Program ASME CODE CLASS:
Class 1,2, and 3 ASME SECTION XI REOUIREMENTS:
ASME Section XI,1989 Edition, Tables IWB-2500-1, IWC-2500-1 and IWD-2500-1 require the performance of a VT-2 examination during the specified pressure tests. IWA-2300 requires that personnel performing the VT-2 examinations be qualified by the owner or the owner's agent in accordance with owners qualification program having levels of competency comparable to SNT-TC-1 A as defined in ANSI N45.2.6.
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(i) Detroit Edison is requesting relief from ASME Section XI requirements to certify VT-2 examiners in accordance with IWA-2300. Detroit Edison is proposing to use the alternatives specified in ASME Code Case N-546 (copy attached). This will eliminate the need to qualify VT-2 examination personnel in the same manner as NDE personnel.
VT-2 requires no special knowledge of technical principles; it is simply an inspection for evidence ofleakage. No special skills or technical training are required in order to observe water dripping from a component or bubbles forming on a surface wetted with a leak detection solution. Therefore, qualification in accordance with the provisions of the Code Case will not present any reduction in quality or safety. In fact, it will facilitate the qualification of those personnel most familiar with the walkdown of plant systems.
The Code Case is ASME approved indicating the ASME Code Committee members reached a consensus that the alternative provides essentially equivalent results to the requirements ofIWA-2300. Detroit Edison agrees with the Code Committee that use of the alternative described in this Code Case will provide an acceptable level of quality and safety.
ALTERNATIVE:
Code Case N-546 provides the following attemative qualification rules for personnel such as licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.
ISI NDE Program Revision 2 Change 0 Part A Page 26 (a) The individual must have at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> plant walkdown experience such as that gained by licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.
(b) At least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of training on Section XI requirements and plant specific procedures for VT-2 visual examination will be completed.
(c) Vision test requirements ofIWA-2231 (1995 Edition) will be satisfied.
In addition, the following actions will ensure consistent quality in the performance of examinations.
- 1. Records of the training and qualifications specified in Code Case N-546 will be provided and maintained in accordance with the Fermi written practice.
- 2. Visual examination will be conducted in accordance with specific writien procedures.
- 3. Visual examination procedures will provide for a documented independent review and evaluation of test results.
APPLICAHLE TIME PERIOD:
Reliefis requested for the second 10-year inspection interval.
l
CASE N-546 CASEE Or ASME DO!!DL AMD Aw VEEEEL Cnnt ISI NDE Program Revision 2 Part A ag 2
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angt any tenNmnenorr esens.
Case N 546 accontance with die referenced standard ti.e.. ANS1 Alternative Respdresumats for Qu= Nae =* ion of N45.2.6, ASNT SNT.TC 1 A. or ASNT CP 180 nu YT 2 Esassdsisties h i
Sect k E N I vided the esannastion personnel are qualined in =ecord.
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ance wnh the following requirements.
(a) At least 40 hr plant walkdown expenence. such Jnguirp What ahoraative to the requirernents of as that passed by licensed and nonlicensed opersters.
IWA 2300 rusy be ased for quallfleason of VT 2 visual M '"'
'F"' " '"I'" "'5' 8"" '"'P' examination personnel?
tion and nondestructive czamination personnel.
f6) At least 4 hrs a( training on Secuon XI recusse.
sments and plant spee:Ge pmeedures for VT 2 mual Acr:fr: la is the opinion of the Committee that YT-
.2 visual esarnisatiosupersonnel need not be qualised (c) Vision ins: seguirements of IWA.2321.10%
nur certified to comparable levels of competence in Edinos.
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r ISI NDE Program Revision 2 Change 0 Part A Page 28 SECOND INTERVAL RELIEF REQUEST RR-A19 COMPONENT FUNCTION / DESCRIPTION:
Insulated portions of High Pressure Coolant Injection (HPCI) turbine / exhaust lines, and associated vents and drains.
)
SYSTEM:
l liigh Pressure Coolant Injection (E41)
ASME CODE CLASS:
Class 2 ASME SECTION XI REOUIREMENTS:
ASME Section XI 1989, IWA-5213(d) (Test Condition IIolding Time) and Code Case N-498-1 (Alternative Rules for 10-year liydrostatic Pressure Testing) which is included in the Fermi inservice Inspection Program, requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time after attaining nominal operating pressure conditions for insulated systems.
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(ii) Detroit Edison is requesting relief from ASME Section XI requirements to maintain a 4-hour hold time prior to the visual examination for the pressure test described in this relief request. Fermi proposes to perform the test using an alternative hold time of 20 minutes. This alternative is necessary because the 4-hour hold time could result in system conditions outside of Technical Specification operating limits.
As part of the Emergency Core Cooling System (ECCS), the IIPCI system is not required to operate during normal plant operation. liowever, the system is periodically tested in accordance applicable inservice testing and Technical Specification requirements. These periodic tests are conducted to verify the operability of system components. The quarterly operability test (24.202.001) normally includes about 30 minutes of pump run time. In order to satisfy ASME Section XI hold time requirement, the test would require a 11PCI pump run for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (hold time plus exam time). Running the IIPCI pump for this duration is not practical and represents an undue hardship on the facility without a compensating increase in the level of quality and safety.
Operating the liPCI pump for this amount of time would subject the facility to excessive heat loads. Control of these heat loads would require the operation of additional ECCS subsystems to remove heat from the suppression pool.
L
[
ISI NDE Program Revision 2 Change 0 Part A Page 29 l
Extended operation of the IIPCI pump would also challenge the Technical Specification l
limitation on maximum suppression pool (torus) water temperature. The Fermi Technical l
Specifications require the torus average water temperature to be maintained less than 105" F during testing which adds heat to the torus. Operating the HPCI pump for a period substantially longer than the system operability test could cause this temperature to be exceeded. If the torus average water temperature exceeds 110*, Technical Specifications require the reactor mode switch to be placed in the shutdown position.
Removal of the insulation from the subject components in order to use the ten minute hold time allowed by the Code or Code Case N-498-1, would be equally burdensome. The impacts associated with insulation removal and reinstallation, include personnel radiation c'.posure,
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radwaste generation, and limited manpower resources are not justified by a compensating increase in the level of quality and safety.
Performing a HPCI system hydrostatic test per IWA 5213 (d) would also be burdensome. A hydrostatic test would require installation of blank flanges and temporary pipe supports, and gagging or removal of relief valves. System out of service time, and radiation exposure incurred in carrying out a hydrostatic test would result in a hardship without a compensating increase in the level of quality and safety.
Other inspection and test activities performed that serve to verify continued system integrity include the following:
i Quarterly inservice testing ofIIPCI raises the pressure of the system to nominal operating conditions. Any leakage would migrate through the insulation over a period of time and would become evident.
Nondestructive examination of circumferential welds per Section XI Table IWB-2500-1, Category C-F-2. The weld selections on this line were random selections because none of the j
welds exceeded the moderate or high stress criteria.
Every 18 months this line is inspected in accordance with the Fermi Leakage Reduction Program per Technical Specification requirements.
ALTERNATIVE EXAMINATION:
The system pressure test described in Code Case N-498-1 will be conducted as required, except that a 20 minute hold time will be used in lieu of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time requirement. The 20 i
minute hold time will allow time for abnormal leaks to migrate through the insulation without challenging the Technical Specification limitation on maximum torus water temperature. Any evidence of abnormal leakage will be investigated by locally removing insulation. A similar alternative for test performance was approved at another nuclear utilities (e.g., Ilope Creek).
I L
I'
(
Retfision 2 Change 0 l
Part A Page 30 Reasonable assurance of system structural integrity is maintained through implementation of the alternative test and by the extent and frequency of other Technical Specification /ASME required system operability tests.
l APPLICABLE TIME PERIOD:
Reliefis requested for the second 10-year inspection interval.
l l
i l
l.-
ISI NDE Program Revision 2 Change O Part A Page 31 l
SECOND INTERVAL RELIEF REQUEST RR-A23 j
COMPONENT FUNCTION / DESCRIPTION:
Pressure Retaining Piping Welds (see attached Table for identification numbers)
SYSTEMS:
l Reactor Recirculation (B31) l Residual IIcat Removal (El1) i Feedwater (N21)
ASME CODE CLASS:
Class 1 ASME SECTION XI REOUIREMENTS:
i ASME Section XI,1989 Edition, Subsection IWB, Table IWB-2500-1, Category B.I, item 9.11 requires a volumetric and surface examination of circumferential piping weUs greater than or equal to 4" diameter. Note 3 of Table IWB-2500-1 requires that the examinathn include essentially 100% of the weld length and volume specified in figure IWB-2500-3.
l BASIS FOR REI,IEF:
l During the course ofinservice examination,4 of 156 Category B-J circumferential welds have been encountered that are impractical to fully examine in accordance with ASME Section XI
(> 90% oflength and volume). Pursuant to 10 CFR 50.55a(g)(5)(iii) Detroit Edison is requesting relief from ASME Section XI requirements to perform complete examinations oflisted piping welds, as described above.
l Fermi proposes to examine these welds to the extent practical within the limits of design and accessibility. Reasonable assurance of piping system structural integrity is provided by the i
Fermi ISI NDE Program as detailed in this relief request. Detroit Edison considers the proposed alternative examination to provide an acceptable level of quality and safety.
l The adjacent weld, which is also a moderate stress weld, is fully examined. Inspections completed through the sixth refueling outage (RF06) have detected no reportable service induced defects in any carbon steel piping welds subject to ISI.
ALTERNATIVE EXAMINATION:
Partial examination of each weld to the greatest extent possible using appropriate surface and ultrasonic examination methods. Additionally, leakage inspections performed at the completion of each refueling outage per Category B-P include all of these welds.
U t
ISI NDE Program Revision 2 Change 0 Part A Page 32 The extent of partial examination and technicaljustification for each is provided below:
Reactor Recirculation (B31)
Category Weld Percentage Limitation Alternative
/ Item Identification Complete Description Examination B-J/B9.11 F W-RS-2-A5 86% PT Pump Insulation Examine
>90% UT Support Ring &
accessible area Brackets This stainless steel weld is a low stress random selection. The weld was given an IGSCC mitigation treatment (Induction Heat Stress Improvement) as defimed in NUREG 0313 Rev. 2, prior to service. Fermi has also implemented an augmented inspection program in accordance with Generic Letter 88-01. The combined Code and GL-88-01 selections result in greater than 50% of all Reactor Recirculation System welds being inspected each interval. The inspection sample set is sufficiently large to provide for reliable detection of representative degradation.
There is no decrease in the ability to detect system degradation as a result of this limitation. Re-designing or removing the obstructions to marginally increase coverage of this weld is impractical. It would also substantially increase man-hours and radiation dose without a compensating increase in plant safety. Detroit Edison believes this altemative provides an acceptable level of quality and safety.
RIlR (Ell)
Category Weld Percentage Limitation Alternative
/ Item Identification Complete Description Examination B-J/B9.11 FW El1-2299-
>50% UT Tee Configuration Examine 0W1 100% PT Limits UT Only accessible area This stainless steel tee-to-pipe weld is a high stress weld selection. The weld was radiographed during construction and sa:isfied Section 111 acceptance criteria. There are also six other high stress locations in the RHR system that were fully examined. The surface of the weld is fully accessible for liquid penetrant examination. Ultrasonic examination is limited to effective scanning from the pipe side only because of reducing-tee configuration. The ultrasonic examination covers all of the base material on the pipe side of the weld and the weld root area.
Because the examination covers the weld root area, which is also the thinnest section of this i
f pipe-to-tee weld zone, there is adequate assurance that IGSCC or fatigue or cracking could be l
detected. Altering the weld design to increase exam coverage would be impractical.
f ISI NDE Program Revision 2 Change 0 Pan A Page 33 Additionally, two adjacent welds on both sides of this weld are fully examined. Fermi has also l
implemented an augmented ia3pection program in accordance with Generic Letter 88-01. The l
combined Code and Generic Letter 88-01 selections result in greater than 50% of all susceptible welds being inspected each interval. The inspection sample set is sufficiently large to provide for reliable detection of representative degradation. There is no decrease in the ability to detect system degradation as a result of this limitation.
Radiographic examination was considered as an alternative but has the following limitations.
The radiation emitted from the pipe would negatively impact the sensitivity of the examination.
Performance of the examination would take approximately one shift to complete and prevent other outage activities from be performed during the radiography evolution. Radiographic examination of the weld would require draining of the recirculation loop piping and a portion of RHR. This would require rduggingjet-pumps and recirc suction lines inside the vessel. RiiR Shutdown cooling would t t be available to remove decay heat. For these reasons radiography is not a feasible altemative for the ultrasonic examination.
Because of the acceptable initial condition, pressure test history and continued performance, the capability to complete the surface exam and gremer than 50 percent of the exam volume including the root area, it is reasonable to conclude there is no significant impact on the level of plant quality and safety by the reduction in volumetric coverage of this weld. Detroit Edison believes this alternative provides an acceptable level of quality and safety.
I l
Feedwater (N21)
Category Weld Percentage Limitation Alternative
/ Item Identification Complete Description Examination B-J/B9.11 FW-N21-2336-
~76% UT Tee to Valve Examine 0W1 100% MT Configuration.
accessible area This carbon steel tee-to-pipe weld is a moderate stress weld selection category as defined in the Fermi UFSAR. The moderate stress category results in an inspection sample of 28% of all Category B9.11 circumferential welds. The increased inspection sample is comp;ised of welds with the highest probability of failure and results in added assurance of system integrity. This is a more conservative approach to selecting welds than a supplemental random selection to bring the examination sample to 25%, as specified in the Code. The inspection sample set exceeds ASME Code requirements and is sufficiently large to provide for reliable detection of system degradation.
l The weld was radiographed during construction and satisfied Section III acceptance criteria. The valve body and weld ends were also radiographed in accordance with NB 2570. The surface of
ISI NDE Program Revision 2 Change 0 Part A Page 34 the weld is fully accessible for magnetic particle examination. Ultrasonic examination is limited because of tee-to-valve configuration. The ultrasonic examination does cover the weld and the weld root area in at least one direction. The base material on the valve side is not fully covered in two directions. Altering the weld design to marginally increase coverage is impractical.
Because of the acceptable initial condition, pressure test history and continued performance, the capability to complete the surface exam and approximately 75% of the exam volume including the root area, it is reasonable to conclude there is no significant impact on the level of plant quality and safety by the reduction in volumetric coverage of this weld. Because the inspection sample population exceeds ASME Code requirements, there is no decrease in the ability to detect system degradafon as a result of this limitation. Detroit Edison believes this alternative provides an acceptable level of quality and safety.
Feedwater (N21)
Category W eld Percentage Limitation Alternative
/ Item Identification Complete Description Examination B-J/B9.11 FW-N21-2336-50% UT Sweepolet to Valve Examine 1WO3 100% MT Configuration accessible area This carbon steel reducer-to-valve weld is a high stress weld selection. The weld was radiographed during construction and satisfied Section 11I acceptance criteria. The valve body and weld ends were also radiographed in accordance with NB 2570. There are also eleven other high stress locations (includes terminal ends) in the Feedwater System that will be fully examined. The surface of the weld is fully accessible for magnetic particle examination.
Ultrasonic examination is limited to effective scanning from the crown of the weld. The ultrasonic examination covers most of the base material on both sides of the weld in one direction. The entire weld and root was scanned in the circumferential direction. Additionally, the high stress weld directly adjacent to this weld was fully examined.
There are over 50 high stress carbon steel weld selections spread among the systems subject to inservice inspection. The Fermi Class 1 inspection population for all systems exceeds ASME Code requirements by 15 welds because moderate stress welds are included in the selection basis. The welds that were selected are the most probable locations for stress related failure. The selection methodology used was more stringent than required by Code. Because of the selection methodology and sample size there is no reduction in capability to detect system degradation as compared to Code requirements. Through the sixth refueling outage (RF06) there were no service induced defects detected. Industry experience does not indicate cracking of carbon steel butt welds to be a problem. All of these reasons indicate that it is impractical to alter the weld design to increase exam coverage for this weld.
L
ISI NDE Program Revision 2 Change 0 l
Part A Page 35 Radiographic examination was considered as an alternative but is undesirable for the following reasons. Draining the feedwater line to perform the examination would make reactor water clean up unavailable and would negatively impact reactor vessel clarity potentially affecting refueling and inspection activitics. It would also prevent drywell and steam tunnel outage activities from be performed during the radiography evolution adding critical path time to the outage schedule.
The benefit ofincreasing the coverage of this weld by radiographic examination has only a small l
potential ofincreasing plant safety margin and a disproportionate impact on other plant activities.
Because of these impacts and since the Fermi inspection program exceeds ASME Code requirements for the sampling program this alternative is not considered to be practical.
Because of the acceptable initial condition, pressure test history and continued performance, the capability to complete the surface exam and approximately 50 percent of the Code exam volume, it is reasonable to conclude there is no significant impact on the level of plant quality and safety by the reduction in volumetric coverage of this weld. Detroit Edison believes this alternative 1
provides an acceptable level of quality and safety.
APPLICABLE TIME PERIOD:
Reliefis requested for the second 10-year inspection interval.
I l
i i
1 ll.
ISI NDE Program Revision 2 Change O Part A Page 36 SECOND INTERVAL RELIEF REQUEST RR-A25 COh1PONENT FUNCTION / DESCRIPTION:
Pressure Retaining Reactor Pressure Vessel (RPV) Circumferential Shell Welds (Welds 4-308A,4-308B,1-313, and 9-307; ref Figure 1)
SYSTEht:
Reactor (B11)
ASN1E CODE CLASS:
Class 1 ASNIE SECTION XI REOUIREMENTS:
ASME Section XI,1989 Edition, Subsection IWB, Table IWB 2500-1, Examination Category B-A, Item No. Bl.11, and the augmented examination requirement of 10CFR50.55a(g)(6)(ii)(A)(2) requires volumetric examination of essentially 100% of RPV circumferential weld and base material regions in the reactor pressure vessel each inspection interval.
HASIS FOR ALTERNATIVE:
Pursuant to 10CFR55.55a(a)(3)(i), and consistent with information contained in NRC Generic letter 98-05, Detroit Edison is requesting an alternative from ASME Section XI requirements to examine essentially 100% of accessible Category B-A circumferential welds and is proposing permanent relief (for the remaining port on of the initial license period) from these examinations.
i The basis for this request for inspection reliefis documented in the report "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," that was transmitted to the NRC in September 1995. The BWRVIP-05 report provides the technical basis for eliminating inspection of BWR RPV circumferential shell welds.
The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the longitudinal shell welds.
The NRC staff has conducted an independent risk-informed assessment of the analysis contained in BWRVIP-05. This assessment also concluded that the probability of failure of the BWR RPV circumfeiential welds is orders of magnitude lower than that of the longitudinal shell welds.
Additionally, the NRC assessment demonstrated that inspection of BWR RPV circumferential welds does not measurably affect the probability of failure.
As discussed during the ACRS Full Committee meeting on July 9,1998 the Staff has completed its evaluation of the BWR Vessel and Internals Project (BWRVIP) recommendations for reduced inspections of the reactor pressure vessel shell welds as described in the BWRVIP-05 report.
Based on the Staff's review, it has been concluded that inservice inspection (ISI) of the BWR RPV circumferential welds is not necessary during the current license term since these welds
151 NDE Program Revision 2 ChangeO Pan A Page 37 have low failure frequencies. The NRC issued a Final Safety Evaluation documenting acceptance of the BWRVIP-05 report on July 28,1998.
The NRC Staffissued Generic Letter 98-05 regarding the use of the BWRVIP-05 report as the basis for BWR licensees to request relief from the requirements to conduct volumetric examinations of the BWR RPV circumferential welds. This independent NRC assessment utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are: the neutron fluence was estimated to be end-of-license mean fluence, the chemistry values are mean values based on vessels types and the potential for beyond design basis events is considered.
Although BWRVIP-05 provides the technical basis supporting this relief request, the following information is provided to show the conservatism of the NRC analysis for the Fermi 2 Nuclear Power Plant. For plants with RPVs fabricated by Combustion Engineering the mean end-of-license neutron fluence use in the NRC PFM analysis was 20 x 10" n/cm. However, at Fermi 2 2
the highest fluence anticipated at the end of the requested relief period (end of the initial license period)is 6.5 x 10" n/cm. Thus, embrittlement due to fluence effects is much lower, and the 2
NRC analysis is conservative for Fermi 2 in this regard. Therefore, there is significant conservatism in the already low circumferential weld failure probabilities as related to Fermi 2.
Other Fermi 2 RPV shell weld information that the NRC staff has requested (GL 99-05) be included in requests for reliefis provided in attached Table 1.
At an August 8,1997 meeting with industry, the NRC staffindicated that the potential for, and consequences of, nondesign basis events (not addressed in the BWRVIP-05 report) should be considered. In particular, the NRC staff stated that nondesign basis cold over-pressure transients should be considered. It is highly unlikely that a BWR would experience a cold overpressure transient. For a BWR to experience such an event multiple operator errors would be required.
At the August 8,1997 meeting, the NRC staff described several types of events that could be precursors to BWR RPV cold over pressure transients. These were identified as precursors because no cold overpressure event has occurred at an U.S. BWR. Also at the August 8 meeting, the NRC staffidentified one actual cold overpressure event that occurred during shutdown at a non-U.S. BWR. This event apparently included several operator errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79 F to 88 F.
As provided in the following discussion, Fermi 2 has in place procedures and Technical Specifications which monitor and control reactor pressure, temperature, and water inventory during all aspects of cold shutdown which would minimize the likelihood of a Low Temperature Over-Pressurization (LTOP) event from occurring. Additionally, these procedures are reinforced through operator training.
The Pressure Test procedures, which are used at Fermi 2, have sufficient procedural guidance to prevent a cold, over-pressurization event. Pressure testing is performed at the conclusion of each outage. The system leakage tests include requirements for operations management to perform a
ISI NDE Program Revision 2 ChangeO Part A Page 37 have low failure frequencies. The NRC issued a Final Safety Evaluation documenting acceptance of the BWRVIP-05 report on July 28,1993.
The NRC Staffissued Generic Letter 98-05 regarding the use of the BWRVIP-05 report as the basis for BWR licensees to request relief from the requirements to conduct volumetric examinations of the BWR RPV circumferential welds. This independent NRC assessment utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are: the neutron fluence was estimated to be end-of-license mean fluence, the chemistry values are mean values based on vessels types and the potential for beyond design basis events is considered.
Although BWRVIP-05 provides the technical basis supporting this relief request, the following information is provided to show the conservatism of the NRC analysis for the Fermi 2 Nuclear Power Plant. For plants with RPVs fabricated by Combustion Engineering the mean end-of-license neutron fluence use in the NRC PFM analysis was 20 x 10" n/cm. However, at Fermi 2 2
the highest fluence anticipated at the end of the requested relief period (end of the initiallicense period) is 6.5 x 10" n/cm. Thus, embrittlement due to fluence effects is much lower, and the 2
NRC analysis is conservative for Fermi 2 in this regard. Therefore, there is significant conservatism in the already low circumferential weld failure probabilities as related to Fermi 2.
Other Fenni 2 RPV shell weld information that the NRC staff has requested (GL 99-05) be included in requests for reliefis provided in attached Table 1.
At an August 8,1997 meeting with industry, the NRC staffindicated that the potential for, and consequences of, nondesign basis events (not addressed in the BWRVIP-05 report) should be considered. In particular, the NRC staff stated that nondesign basis cold over-pressure transients should be considered. It is highly unlikely that a BWR would experience a cold overpressure transient. For a BWR to experience such an event multiple operator errors would be required.
At the August 8,1997 meeting, the NRC staff described several types of events that could be precursors to BWR RPV cold over pressure transients. These were identified as precursors because no cold overpressure event has occurred at an U.S. BWR. Also at the August 8 meeting, the NRC staff identified one actual cold overpressure event that occurred during shutdown at a non-U.S. BWR. This event apparently included several operator errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79"F to 88 F.
As provided in the following discussion, Fermi 2 has in place procedures and Technical Specifications which monitor and control reactor pressure, temperature, and water inventory during all aspects of cold shutdown which would minimize the likelihood of a Low Temperature Over-Pressurization (LTOP) event from occurring. Additionally, these procedures are reinforced through operator training.
l The Pressure Test procedures, which are used at Fermi 2, have sufficient procedural guidance to prevent a cold, over-pressurization event. Pressure testing is performed at the conclusion of each outage. The system leakage tests include requirements for operations management to perform a
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Change O Part A Page 38
" pre-job briefing" with all essential personnel. This briefing details the anticipated testing evolution with special emphasis on: conservative decision making, plant safety awareness, lessons learned from similar in-house or industry operating experiences, the importance of open communications, and finally, the process in which the test would be aborted if plant systems responded in an adverse manner. Vessel temperature and pressure are required to be monitored throughout these tests to ensure compliance with the Technical Specification pressure-temperature curve.
Additionally, to ensure a controlled, deliberate pressure increase, the rate of pressure increase is administratively limited throughout the performance of the test. If the pressurization rate exceeds this limit, direction is provided to remove the CRD pumps which are used for pressurization, from service.
With regard to inadvertent system injection resulting in an LTOP condition, the high pressure make-up systems (High Pressure Coolant injection (IIPCI) and Reactor Core Isolation Cooling (RCIC) systems, as well as the normal feedwater supply (via the Reactor Feedwater Pumps) at Fermi 2 are all steam driven. During reactor cold shutdown conditions, no reactor steam is available for the operation of these systems. Therefore,it is not possible for these systems to contribute to an over-pressure event while the unit is in cold shutdown.
The Standby Feed Water (SBFW) system is an available high pressure electric driven make up system. The SBFW system does not automatically inject water into the RPV. The SBFW system requires deliberate operator action to open the injection isolation valve. Procedures are in place to administratively control the use of the SBFW system.
In the case oflow pressure system initiation, the Fermi 2 pressure-temperature limit curves for hydrostatic testing as provided in Fermi Technical Specifications, permit pressures up to 312 psig at temperatures from 71 F up to 100"F. Above 100 F, the permissible pressure increases immediately to near 600 psig and increases rapidly with increasing temperature. The shutoff head for the Core Spray and Residual lleat Removal Pumps are both below 400 psig. Therefore, the potential for an over-pressurization event which would exceed the pressure-temperature limits, due to an inadvertent actuation of this system is very low.
Procedural control is also in place to respond to an unexpected or unexplained rise in reactor water level, which could result from a spurious actuation of an injection system. Actions specified in this procedure included preventing condensate pump injection, securing ECCS system injection, tripping CRD pumps, terminating other injection sources, lowering RPV level via the RWCU system, and the steam line drains.
In addition to procedural barriers, Licensed Operator Training is given which further reduces the possibility of the occurrence of LTOP cvents. During Initial Licensed Operator Training the following topics are covered: Brittle fracture and vessel thermal stress; Operational Transient (OT) procedures, including the OT on reactor high level: Technical Specifications training, t
ISI NDE Program Revision 2 Change O Part A Page 39 including discussion of Pressure / Temperature (P/T) Limits; and Simulator Training of plant heatup and cooldown including performance of surveillance tests which ensure pressure-temperature curve compliance.
L1 addition to the above, continuous review of industry operating plant experiences is conducted to ensure that the Fermi 2 procedures consider the impact of actual events, including potential LTOP events. Appropriate adjustments to the procedure.s and associated training are then implemented to preclude similar situations from occurring at Fermi 2.
Based on the above, the probability of a cold over-pressure transient is considered to be highly unlikely.
The NRC staff transmitted a Request for Additional Information (RAI) regarding the BWRVIP-05 report to the BWR Vessel and Internals Project (BWRVIP). The BWRVIP provided a response to the RAI that included additional information on the BWRVIP PFM analysis, comparisons to the NRC Staff PFM analysis, and additional information regarding beyond design basis cold overpressure transients. We believe the BWRVIP-05 report and the NRC Final Safety Evaluation Report analysis provide sufficient basis to support this relief request.
Based on the documentation in BWRVIP-05, the risk-informed independent assessment performed by the NRC staff and the discussion above, permanent relief (for the remaining ponion of the initial license period) from completing inspection of the RPV circumferential shell
)
welds at Fermi 2 isjustified.
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ISI NDE Program Revision 2 ChangeO Part A Page 40 TABLE 1 Fermi 2 RPV Shell Weld Information Bounding Circumferential Weld Neutron fluence at the end of the 6.5 x 10" n/cm2 requested relief period (upper bound value)
Initial (unirradiated) reference temperature
-50"F l
Weld Chemistry factor (CF) 236 F Weld Copper content 0.23 %
j Increase in reference temperature due to 31.6 F irradiation (ARhn)
Margin term 31.6 F Mean adjusted reference temperature (ART)
-18.4 F Upper bound adjusted reference temperature (ART) 13.2 F ALTERNATIVE:
The beltline circumferential weld (1-313) was partially examined during the first inspection interval (approximately 54% complete, RF02, Spring 1991). Additionally, Detroit Edison will perform examination of approximately 5% of the Fermi 2 RPV circumferential weld areas only at the intersection of longitudinal seams.
APPLICABLE TIME PERIOD:
Reliefis requested for the remaining portion of the initial license period..
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1 ISI NDE Program Revision 2 Change 0 Part A Page 42 SECOND INTERVAL RELIEF REQUEST RR-A26 COMPONENT FUNCTION / DESCRIPTION:
Extension of Containment Piping Welds SYSTEMS:
El1-Residual Ileat Removal E21-Core Spray E41-High Pressure Coolant Injection E51-Reactor Core Isolation Cooling G51 Torus Water Management P11-Demineralized Service Water i
P44-Emergency Equipment Cooling Water
)
T46-Standby Gas Treatment T48-Containment Atmosphere Control ASME CODE CLASS:
Class 2 ASME SECTION XI REOUIREMENTS:
ASME Section XI,1989 Edition, Subsection IWE-1220(d) provides an exemption from IWE required examinations for piping that is part of the containment system or which penetrates or is attached to the containment vessel. The exemption subsequently requires this piping to be examined in accordance with IWB or IWC as appropriate.
l BASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(i) Detroit Edison is requesting relief from ASME Section XI requirements to perform the surface or volumetric examinations specified in the 1989 Edition of ASME Section XI for piping classified as extensions of containment. The proposed alternative of visual examination is consistent with the examination of other containment items that only require visual examination.
Detroit Edison has identified a new subset of piping which is considered an extension of containment (penetrates containment and within the outboard isolation valve) where the only reason for its selection is the containment function. This is because the system function is either not safety related (e.g., RIIR containment / suppression pool spray lines) or that if the rules of IWC were applied, the piping would be exempt from examination per IWC 1220. This is because the piping is either open ended beyond the last shut off valve or the line process conditions are less than or equal to 275 psig and at a temperature equal to or less than 200 F.
Since the piping selection is based solely on the containment function, it would not make sense to exempt the piping based on the lack of a safety related system /line function or configuration and design parameters of the process stream.
ISI NDE Program Revi bn 2 Change 0 Pan A Page 43 l
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If the selected piping was subject to IWE requirements for Category E-B, Item E3.10 (containment penetration welds) the examination method would be a visual examination.
l Additionally,10 CFR 55a(b)(2)(x)(c) has made containment weld inspection optional because J
there has been no degradation mechanism specific to containment welds. Application of the IWC rules for Category C-F-2 as specified in IWE-1220(d) would result in surface and volumetric examination as required depending on the nominal pipe wall thickness. As stated l
previously, considering only the IWC selection requirements, the piping could be exempted from examination. Since the only reason for selecting the subject piping is the containment function, it seems appropriate to apply the IWE inspection methodology rather than the IWC inspection methodology.
l ALTERNATIVE EXAMINATION:
Detroit Edison proposes to perform a visual examination (VT-1) of all selected welds, except those selected based on the high or moderate stress categories defined in UFSAR 5.2.8.8, that will be examined as specified by Table IWC-2500-1 for category C-F. The sample size will be at least 7.5% of the total number ofpressure retaining extension of containment welds subject to i
)
examination requirements. This percentage meets the 1989 Section XI selection rate requirements for Category C-F-2 pressure retaining welds. This alternative is equivalent to the
.j IWE methodology for examination of penetration welds.
In addition to the visual examination, the extension of containment piping will be subject to 10 CFR 50 Appendix J leakage rate testing.
4 APPLICABLE TIME PERIOD:
(
Reliefis requested for the second 10-year inspection interval.
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ISI NDE Program Revision 2 Change O Part A Page 44 SECOND INTERVAL RELIEF REQUEST l
RR-A27
_ COMPONENT FUNCTION / DESCRIPTION:
Pressure Retaining Closure IIead Nuts SYSTEM:
Reactor (B11)
A.SME CODE CLASS:
Class 1 ASME SECTION XI REOUIREMENTS:
ASME Section XI 1989, Table IWB 2500-1, Examination Category B-G-1, item No. B6.10 (Closure Ilead Nuts) requires a surface examination of all reactor vessel nuts.
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(g)(5)(iii) Detroit Edison is requesting relief from ASME Section XI requirements to examine the surface of the reactor vessel nuts, because within the limits of design it is impractical to do so.
Table IWB 2500-1 specifies the surface exam method of examination but the examination figure has never been prepared.
The closure head nut configuration does not allow for an effective magnetic particle examination for service related defects. The MT method requires two-directional coverage to detect surface flaws. Only the external surface is capable of being satisfactorily examined in two directions.
The threaded area could be examined for detection of radial reflectors using a central conductor.
Ilowever service related planar defects would more likely be orientated circumferentially in the threads which would make effective examination impractical. Since the nuts receive a compressive load from the stud it is extremely unlikely that there would be any degradation on the external surface other than mechanical damage caused by installation and removal or corrosion. All RPV nuts received a surface examination for manufacturing defects by the material manufacturer.
Liquid penetrant examination is not practical because of the difficulty that would be associated with removing thread lubricant.
ALTERNATIVES:
Perform a VT-1 examination of all visible surfaces of the RPV nuts.
APPLICAHLE TIME PERIOD Reliefis requested for the second 10-year inspection interval.
ISI NDE Program Revision 2 Change O Part A Page 45 SECOND INTERVAL RELIEF REQUEST RR-A28 COMPONENT FUNCTION / DESCRIPTION:
Pressure Retaining Bolted Connection Leakage SYSTEM:
All Systems included in the ISI NDE Program Plan ASME CODE CLASS:
Class 1,2, and 3 ASME SECTION XI REOUIREMENTS:
l ASME Section XI 1989, IWA 5250(a)(2) requires the following corrective measures ifleakage is observed during VT-2 examination during the system pressure test at bolted mechanical joints; 1) remove all the bolting material associated with thatjoint,2) perform a VT-3 examination for corrosion and 3) evaluate the conditions in accordance with IWA-3100.
l BASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(ii) Detroit Edison is requesting relief from ASME Section XI requirements to remove bolting for visual examination when leakage is noted at a bolted connections. This request for reliefis based on the hardship associated with removal without a compensating increase in the level of quality or safety.
Fermi 2 is a boiling water reactor (BWR) and the reactor coolant system and associated systems do not experience the corrosive environment from boric acid residues, as would a pressurized water reactor (PWR). Therefore, there is no reason to suspect degradation of bolting caused solely by the chemistry ofleaking coolant.
l The purpose ofIWA-5250(a)(2) is to determine ifinservice leakage has degraded the bolting l
material. Therefore bolting that is new or was visually examined during joint disassembly would not warrant removal. Additionally, bolting that is in air or gas service should also be excluded.
i Bolting such as control rod drive (CRD) flange cap screws have a history ofleaking upon return to service but decreases over time. This bolting is a chrome alloy material that is resistant to general corrosion. CRDs are rebuilt periodically and bciting is VT-1 examined and reinstalled or replaced as necessary.
Bolting in flanged joints are frequently visible because of the space between the flanges. And while flange or valve bonnet leakage is not desirable the prudent corrective measure may be to verify torque and re-tighten bolting as necessary.
The 1989 Code is too restrictive and does not allow for evaluation and application of prudent engineeringjudgement. Satisfying the Code requirement for removing bolting may require significant planning and scheduling due to operational concems and personnel safety. In cases of l
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1 ISI NDE IYogram Revision 2 Change 0 Pan A Page 46 unisolatable or non-redundant piping, the requirement to remove the bolting in order to conduct the visual examination may necessitate shutdown of the plant and result in unnecessary plant I
transient cycles.
ALTERNATIVES:
Detroit Edison will document an evaluation of the flange leakage and determine the appropriate course of action. The evaluation will consider the potential for bolting degradation as well as the cause of the leakage. The evaluation will determine whetner bolt tightening or removal is necessary to ensurejoint integrity.
Should bolting removal be necessary, Detroit Edison proposes to remove the bolt nectest the leakage source, as specified in the 1990 Addenda of ASME Section XI. If the bolt shows evidence of significant degradation additional bolts for that connection will be removed and examined and evaluated in accordance with IWA-3100.
APPLICAHLE TIME i'ERIOD Reliefis requested for the second 10-year inspection interval.
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ISI NDE Program Revision 2 Change O Part A Page 47 SECOND INTERVAL RELIEF REQUEST RR-A29 COMPONENT FUNCTION / DESCRIPTION:
Austenitic Stainless Steel BWR Coolant Piping Welds SYSTEM:
All Systems included in the ISI NDE Program Plan ASME CODE CLASS:
Class 1 and 2 ASME SECTION XI REOUIREMENTS:
ASME Section XI 1989, IWB-2430 provides requirements for additional examinations when indications are revealed that exceed the acceptance standards ofIWB-3000.
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(i) Detroit Edison is requesting relief from ASME Section XI requirements for additional weld examinations for welds subject to Generic Letter 88-01 and NUREG 0313 Revision 2, because these regulatory documents provide alternative sample expansion guidance that considers Intergranular Stress Corrosion Cracking (IGSCC) susceptibility. This methodology ensures that welds with similar risk, i.e. weld category, pipe size, system, and location, for cracking are examined while maintaining radiation exposure of examination personnel as low as reasonably achievable.
The Code specified expansion methodology only considers Code item numbers and not materials or susceptibility to degradation. For example, the Code item B9.1I would include carbon steel as well as stainless steel welds. The carbon steel welds would not be subject to IGSCC.
Therefore, it would not be appropriate to include those items in the sample expansion.
ALTERNATIVES:
When examinations are being performed to satisfy the requirements of Generic Letter 88-01 and NUREG-0313 in addition to the Code requirements, sample expansion resulting from unacceptable IGSCC flaw indications will be performed using the methodology specified in Generic Letter 88-01 and NUREG-0313.
APPLICABLE TIME PERIOD Reliefis requested for the second 10-year inspection interval.
ISI NDE Program Revision 2 Change O Pan A Page 48 SECOND INTERVAL RELIEF REQUEST RR-A30 COMPONENT FUNCTION / DESCRIPTION:
Pressure Retaining Piping Welds, Categories B-J and C-F-2 SYSTEM:
All Systems inclnded in the ISI NDE Program Plan ASME CODE CLASS:
Class 1 and 2 ASME SECTION XI REOUIREMENTS:
ASME Section XI 1989, Tables IWB-2500 and IWC 2500 require volumetric examination of pressure retaining welds in piping NPS 4 and larger. When ultrasonic examination is applicable, IWA-2232 requires conduct in accordance with Appendix 1. Appendix I specifies that ultrasonic examination ofpiping welds be performed per Appendix Ill.
BASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(i) Detroit Edison is requesting relief from ASME Section XI requirements to use the amplitude based examination techniques described in Appendix 111 because more effective techniques are available.
I The Utility Performance Demonstration Initiative (PDI) developed a program based on the 1992 j
Edition with the 1993 Addenda of ASME Section XI. This program requires that ultrasonic equipment, procedures, and examiners be qualified on flawed and notched specimens with configurations similar to those found in the plant. Consequently, the PDI Program provides a higher degree of reliability for detection and characterization of flaws when compared to the conventional amplitude-based ultrasonic techniques required by the 1989 Edition of ASME Section XI.
The NRC issued a letter to the Chairman of the BWR Owners Group on March 1,1996 discussing the transition from the Intergranular Stress Corrosion Cracking (IGSCC) Qualification Program to the PDI Program for qualification requirements applicable to procedures and personnel for Generic Letter 88-01/NUREG-0313 examinations. The letter stated that personnel qualification and subsequent requalification for the IGSCC Program could be obtained through the PDI Program. The techniques developed and qualified through PDI are recognized as being superior to those specified in Appendix III.
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l ISI NDE Program Revision 2 Change O Part A Page 49 ALTERNATIVES:
When examinations are being performed to satisfy the requirements of Generic Letter 88-01 and NUREG-0313, PDI qualified personnel and procedures will be used. For all other examinations, I
the techniques developed through the PD1 program will be used by certified examination personnel trained in their use.
APPLICABLE TIME PERIOD Reliefis requested for the second 10-year inspection interval.
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J ISI NDE Program Revision 2 Change O Part A Page 50 5.0 Program Support Drawings-5.1 Inservice Inspection Classification Boundary Drawings The systems or portions of systems subject to the examination requirements of the ISI-NDE Program for Fermi 2 and the associated Class 1,2, and 3 boundaries are documented on ISI Classification Boundary Drawings. The ISI Classification Boundary Drawings are listed in Table A-5-5.1.
The system Inservice Inspection classifications, developed specifically to define the extent to which Section XI requirements will be applied, differ somewhat from the ASME Section III design classifications. These differences occur because systems, or portions of systems, have been optionally upgraded in design and because ISI classifications are limited to safety related systems which contain water, steam, or radioactive materials. The guidance provided by NUREG 0800, Regulatory Guide 1.26, and 10CFR50 were used in establishing these boundaries.
5.2 Inservice Inspection (ISI) Isometric Drawings for Class 1 and 2 Components and their Supports The ASME Section XI, Class 1 and 2 isometrics referenced in Table A-5-5.2 of this program identify components subject to inservice inspection. Drawings are updated to conform to as-built configuration of plant piping systems following modifications.
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ISI NDE Program Revis:on 2 Change O Part A Page 51 TABLE A-5-5.1 i
FERMI-2 ISI CLASSIFICATION BOUNDARY DRAWINGS Drawing Title Dwg. No.
PIS No's ISI Code Class Associated with Dwg.
Nuclear Boiler System 6M721-5808-1 B21 1
Main Steam Isolation Valve 6M721-5808-1 B21-06 1
Leakage Control System 2
Main and Reheat Steam 6M721-5808-1 B21 1
Systems 6M721-5822 N11/N30 2
Reactor Recirculation System-6M721-5809 B31 1
)
Nuclear Boiler System Control Rod Drive System 6M721-5810-1 Cl1 2
l 2
CRD Scram Discharge System 6M721-5810-2 Cl1 2
Stand-By-Liquid Control 6M721-5811 C41 1
System 2
Residual lieat Removal 6M721-5813-2 El1 1
(RilR) Division 1 2
Residual lleat Removal 6M721-5813-1 El1 1
(RiiR) Division 11 2
l RIIR Service Water Make Up 6M721-5813-3 El1 3
l Decant and Overflow Systems P45 3
R30 3
)
l Core Spray System 6M721-5814 E21 1
l 2
Ifigh Pressure Coolant 6M721-5815 E41 1
Injection System (11PCI) 2 Reactor Core Isolation 6M721-5816 E51 1
Cooling System (RCIC)
Reactor Water Clean-Up 6M721-5818 G33 1
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ISI NDE Program Retrision 2 ChangeO Part A Page $2 TABLE A-5-5.1 (Cont'd)
FERMI-2 ISI CLASSIFICA110N BOUNDARY DRAWINGS Drawing Title Dwg. No.
PIS ho's Associated ISI Code Class with Dwg.
Fuel Pool Cooling and 6M721-5819 G41 2
Clean-Up System Torus Water Management 6M721-5820 GSI 2
System Feedwater System 6M721-5821 N21 1
Sample Line Tie-In And 6M721-5824 P34 1
Return Post Accident 2
Sampling System Emergency Equipment 6M721-5825-1 P44 3
Cooling Water Division i Emergency Equipment 6M721-5825-2 P44 3
Cooling Water Division 11 (Post Loca) Combustible 6M721-5830-2 T48 2
Gas Control System t.
ISI NDE Program Revision 2 Change 0 Part A Page 53 TABLE A-5-5.2 FERMI-2 ISI ISOMETRIC DRAWINGS Class 1 Drawing Description 6M721-2000-5 Index for Inservice Inspection Isometrics 6M721-2001-5 Legends and Symbols for Inservice Inspection Isometrics 6M721-2192-5 RCIC Steam Line in Drywell 6M721-2297-5 11PCI Steam Line in Drywell 6M721-2298-5 RiiR Return Line Div.1 (North) 6M721-2299-5 RilR From Recire Pump Suction 6M721-2327-5 RI1R Return Line Div. 2 (South) 6M721-3052-5 Core Spray Piping Div.1 (North) 6M721-3053-5 Core Spray Piping Div. 2 (South) 6M721-3096-5 Reactor Water Clean-Up Suction From the RPV Bottom licad Drain 6M721-3536-5 Reactor Feedwater Piping (North) 6M721-3537-5 Reactor Feedwater Piping (South) 6M721-5351-5 Reactor Water Clean-Up Suction From the Recire Loops 6M721-5352-5 Main Steam Loop A 6M721-5353-5 Main Steam Loop B 6M721-5354-5 Main Steam Loop C 6M721-5355-5 Main Steam Loop D 6M721-5356-5 Reactor Recire Loop A Ring lieader 6M721-5357-5 Reactor Recire Loop A Pump Suction and Discharge 6M721-5358-5 Reactor Recire Loop B Ring IIcader 6M721-5359-5 Reactor Recire Loop B Pump Suction and Discharge 6M721-5360-5 Reactor Vessel No. I 6M721-5361-5 Reactor Vessel No. 2 6M721-5362-5 Reactor Vessel No. 3 6M721-5363-5 Reactor Vessel No. 4 6M721-5364-5 Reactor Vessel No. 5 6M721-5365-5 Reactor Recire Pump Class 2 6M721-2095-5 Combustible Gas Control Return IIeader to Torus (Div.1) 6M721-2097-5 Combustible Gas Control Retum IIcader to Torus (Div. 2) 6M721-2979-5 Standby Liquid Control Explosive Valves and Line to Drywell Pen. X-42 6M7E 't035-5 RiiR licad Spray From Retum IIcader to Drywell Penetration 6M'al 5144-5 North Core Spray Pump Discharge to RPV 6M72 M145-5 North Core Spray Min. Low-Flow By-Pass & Test Line 6M721-3146-5 RilR Retum (North) From RiiRIIX to Drywell Pen. No. X-13B 6M721-3147-5 Core Spreay Pump (South) Discharge to RPV Penet.
6M721-3148-5 Core Spray Pump (North) Suction From Supression Chamber
is: NDE Program Revision 2 Change 0 Part A Page 54 TABLE A-5-5.2 Class 2 cont'd 6M721-3149-5 Core Spray Pump (South) Suction From Supression Chamber 6M721-3150-5 South Core Spray Min Low-Flow By-Pass & Test Line 6M721-3151-5 RIIR Return (South) From RIIR IIX to Drywell Pen. No. X-13 A 6M721-3153 South RllT Pump Suction From Suppression Chamber 6M721-3154-5 North RiiT Pump Suction From Suppression Chamber 6M721-3157-5 RilR Pump Discharge North 6M721-3158-5 RIIR Pump (N & S) Discharge to lleat Exchanger 6M721-3159-5 RIIR Containment Spray (North) From lieader to RPV 6M721-3160-5 RIIR Test Line & Suppression Chamber Spray IIcader (North) 6M721-3161-5 RiiR Test Line (South) From 24" IIcader to Suppression Chamber 6M721-3162-5 IIPCI Turbine Exhaust 6M721-3163-5 liPCI Booster Pump Suction From Suppression Chamber & Condensate Storage lleader 6M721-3164-5 RIIR Containment Spray (South) From Return IIdr. To Drywell Pen.
6M721-3167-5 IIPCI Pump Discharge to South Reactor Feedwater lieader 6M721-3169-5 IIPCI From Pump Discharge to Condensate Storage System 6M721-3172-5 IIPCI Steam Supply From RPV to llPCI Turbine Stop Valve 6M721-3177-5 R11R Pump Discharge (South) 6M721-3184-5 RiiR Service Water Supply & Return to IIx's (Div. 2) 6M721-3258-5 Main Steam From Drywell to Turbine Manifold 6M721-3259-5 52" Main Steam Maniforld & Piping in Turbine Bldg.
6M721-3361-5 Standby Liquid Control Pump Suction 6M721-3519-5 RIfR IIead Spray 6M721-3669-5 RiiR Supply IIcader to FPC System 6M721-4611-5 Relief Line From RiiR IlX (South) 6M721-4612-5 Relief Line From RIIR IIX (North) 6M721-5370-5 RilR Div. 2 Ileat Exchanger B 6M721-5371-5 RIIR Div.1 IIcat Exchanger A 6M721-5372-5 Control Rod Drive Scram Discharge Volume "A" North 6M721-5373-5 IIPCI Booster to IIPCI Main Pump 6M721-5374-5 Standby Liquid Control Pump Discharge & Test Tank Line 6M721-5375-5 Control Rod Drive Scram Discharge Volume "B" South 6.0 Inservice Inspection Program (Plan) Tables (NDE)
The accompanying table lists the components or areas that are to be examined during the interval. Listed in an order following the items presented in the ASME Section XI, Subsections IWB, IWC and IWD, the tables contain the following information:
Code Class: is the ASME Section XI Classification as determined in accordance with the Code of Federal Regulations (10CFR50.55a) and the guidance provided in Regulatory Guide 1.26, and NUREG 0800.
Revision 2 Change 0 l
Part A Page 55 Interval: refers to the 120-month inspection interval as discussed in Section 2.0 of this document.
1 Code Category: is the Examination Category as denned by ASME Section XI, Subarticles IWB-2500, or iWC-2500, or IWD-2500.
Item Number: lists the item No as defined by ASME Section XI, Subarticles IWB-2500, IWC-2500, or 1WD-2500.
Item numbers that are not applicable to Fermi 2 design were not included in Table A.
This includes all Code Items specific to PWRs (Pressurizers, Steam Generators, Class 1 Heat Exchangers). Additionally, there are none of the following items applicable to BWRs. Category B-A Reactor Pressure Vessel repair welds, item Bl.150. Category B-F dissimilar metal welds < 4" NPS, items B5.140 and B5.150. Pump casing or valve body welds in ASME Class I systems Category B-L-1 or B-M-1. Pump casing or valve body welds in ASME Class 2 systems Category G-G. Category C-B Nozzles in vessels,0.5" in thickness.
l Identification: Provides the unique identification for each weld or item.
Exam Required: identifies the method of examination proposed to satisfy the Code requirement for volumetric, surface, or visual examination. The specific examination l
selected is there shown for the component i.e. UT, PT, MT, or VT. (see list of j
abbreviations for expanded definitions).
l Examination methods delineated in the tables are intended to be representative of the ISI i
practice to be used or of previous methods utilized. In either case, it should be recognized that either UT or RT are acceptable volumetric exams and either PT or MT are acceptable surface exams. Unique weldjoint parameters may, of course, dictate more restrictive selection criteria; e.g., high background radiation will preclude RT, stainless materials will preclude MT, etc. It is intended that the process which selects exam methods for inspections under this plan treat UT and RT as interchangeable and PT and MT as interchangeable with consideration given to past practice in light of the reproducibility of results.
Relief Request: if applicable, indicates the request for relief applicable in accordance with 10CFR50.55a (g)(5)(iii). See Section 4.0 of this document.
Inspection Period: Indicates the scheduled period that the Code or Augmented examination (regulatory or licensing cemmitment) is to be completed. A period is defined as one of three 1/3 year time periods within the 120-month (10 year) interval when the specific examination is scheduled. A period can vary by +/- 1 year collectively over the 10 year interval. Each period contains the specific outage code indicating that the required exam has either been scheduled or completed.
ISI NDE Program Revision 2 Change 0 Part A Page 56 Set. Basis: shows the abbreviation for the basis for selection of a component for examination.
NOTE A tentative schedule of specific examinations has been provided for the second 10 year interval. All exams are scheduled for inspection in accordance with the rules of ASME Section XI, IWA, IWB, IWC, IWD and IWF, and as augmented by specific commitments (e.g., NUREG 0313 or NUREG 0619). Future revisions to this program (plan) shall be issued to reflect actual examinations to be performed during each refuel outage as well as all examinations completed during previous outages.
Remarks: is reserved for additional information to explain, amplify, or provide added details necessary to clarify the examination requirements.
l 1
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ISI NDE Program Revision 2 Change 0 Part A Page 57 6.2 List of Abbreviations: The following abbreviations are used:
Plant Identification System (PIS)- Codes for Plant Systems 1
i B11 PlS Number for the Reactor Pressure Vessel PIS Number for the Nuclear Boiler System B21 B31 PIS Number for the Reactor Recirculation System Cl1 PIS Number for the Control Rod Drive System C41 PIS Number for the Standby Liquid Control System El1 PlS Number for the Residual lleat Removal System E21 Pts Number for the Core Spray System E41 PlS Number for the liigh Pressure Coolant injection System E51 PIS Number for the Reactor Core Isolation Cooling System Gil PIS Number for the Rad Waste System G33 PIS Number for the Reactor Water Cleanup System G41 PIS Number for the Fuel Pool Cooling System G51 PlS Number for the Torus Water Management System N20 PIS Number for the Condensate System N21 PIS Number for the Feedwater System Pil PIS Number for the Condensate Storage and Transfer N30 PIS Number for the Main Steam System P34 PIS Number for the Post Accident Sampling System PlS Number for the Reactor Building Closed Cooling Water System P42 P44 PlS Number for the Emergency Equipment Cooling Water System P45 PIS Number for the Emergency Equipment Service WaterSystem P50 PIS Number for the Compressed Air System R30 PIS Number for the Emergency Diesel Generator and Service Water System T23 PIS Number for the Containment System T46 PIS Number for the Standby Gas Treatment System T48 PIS Number for the Combustible Gas Control System (T4804)
T49 PIS Number for the Primary Containment Pneumatic Supply System PIS Number for the Primary Containment Monitoring System T50 T71 PIS Number for Instrumentation Acronyms Used to Identify Plant Systems CGC Combustible Gas Control CRD Control Rod Drive Core Spray CS FPC Fuel Pool Cleanup liigh Pressure Coolant injection liPCI Reactor Core isolation Cooling RCIC RiiR Residual lleat Removal RRC Reactor Recirculation l
SDV Scram Discharge Volume Standby Liquid Control SLC l
l l
ISI NDE Program Revision 2 Change O Part A Page 58 Non-Destructive Examination Method Abbreviations Magnetic Particle Examination MT Liquid Penetrant Examination PT UT Ultrasonic Examination VT Visual Examination VT-1
- Visual Examination per IWA-2211 VT-2
- Visual Examination per IWA-2212 VT-3
- Visual Examination per IWA-2213 UT Mech.
- UT Mechanized UT Mech / Man. - UT Mechanized or Manual Weld Selection Hasis Abbreviations liigh Cumulative Usage llCU liigh Stress llS MS Moderate Stress Random selection of structural discontinuity weld R
TE Terminal End i
Augmented A
DM Dissimilar Metal Weld Plant Component and Weld Terminology Abbreviations Control Rod Drive llousing
' CRDil Pipe Expansion Joint EXPJT Flange Bolted Connection FBC FW Field Weld j
llX licat Exchanger Heat Exchanger Shell IIXS Inner Bore Region (Nozzle)
IBR 1RS
- Inner Radius Section (Nozzle) 1111
- Incore Instrumentation liousing Longitudinal Downstream (Seam Weld)
LD Longitudinal Upstream (Seam Weld)
LU Integral Attachment Weld Directly onto the Pressure Boundary of the PAD Pipe Piping Support Field Weld PSFW Primary Steam (Nuclear Steam Supply System)
PS Recirculation Discharge RD RS Recirculation Suction Scram Discharge Volume SDV SW
- Shop Weld lianger Support Welded Directly onto the Pressure Boundary of the Pipe TRUNION Valve Body and Bonnet llousing VBB
Revision 2 Change 0 Pan A Page 59 Generie Miscellaneous Abbreviations Boiling Water Reactor BWR Drawing DWG DW Drywell EF2 Enrico Fermi 2 in.
Inches N/A Not Applicable NUREG
- Nuclear Regulatory Guide l
- Pressurized Water Reactor ReliefRequest RR Reactor Builing RB l
TB Turbine Building RPV Reactor Pressure Vessel Component Support Abbreviations A
Anchor C
Constant Support l
G Guide R
Rigid Support SP
- Spring Hanger Snubber Abbreviations Mech Mechanical Snubber Hyd
- Hydraulic Snubber Outage Codes XX C or S Completed Exam or Scheduled Exam j
Refuel Outage Sequential Number EXAMPLE: 07C - Seventh Refueling Outage, Completed Exam 08S - Eighth Refueling Outage, Scheduled Exam i
i NOTE: Modifications to the above Codes by inclusion of a "A"(Augmented Examination)
"P" for (Partial Inspection) or an "L" for (Limited Examination) will be utilized as
ISI NDE Program Revision 2 Change 0 Part A Page 60 necessary. Generally, the extent of the examination completed or limitations of the examination will be noted in the adjacent remarks column for each component.
Description of Building Floor Abbreviations:
DW-Primary Containment (Drywell)
RBB - Reactor Building Basement /Sub-basement RBI -Reactor Building First Floor RBI -Reactor Building Second Floor RBI -Reactor Building Third Floor TB1 -Turbind Building First Floor TB2-Turbine Building Second Floor TB3 -Turbine Building Third Floor l
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D C N 1
3 ISI-NDE Prograrn Rev.2 Change 0 Table A Page 49 NOTE 1 SECTION XI REQUIREMENTS:
Examination Category B-F, Pressure Retaining Dissimilar Metal Welds, and Examination Category B.J, Pressure Retaining Welds in Piping.
1 l
The code does not define a transition point between the Reacu Pressure Vessel and f
piping components. Specifically, the code does not define whether a vessel nozzle to safe end weld is a piping weld or a reactor pressure vessel weld. In addition, the nozzle to safe end weld is considered a terminal end in accordance with Category B-J Footnote (1)(a).
i To further complicate the situation, examination categories B-F and B-J contain duplicate examination requirements for dissimilar metal pressure retaining welds in piping.
DETROIT EDISON INTERPRETATIONS: Nozzle to safe end welds will be considered pressure retaining piping welds and subject to the examination requirements of Examination Category B-F (Item Nos. B5.10 and B5.20) and Category B-J (Item No.
B9.11 and B9.21).
Dissimilar metal piping welds will be subject to the examination requirements of Examination Categories B-F (Item Nos. B5.130 and B5.140) and B-J (Item No.B9.11 or B9.21).
Since the examination requirements of Examination Categories B-F and B-J are identical for a given size and type of weld, the examinations performed will be used to satisfy the requirements of both Examination Categories.
NOTE 2 Reference Detroit Edison Documents NRC-88-0243, NRC-89-0297, and NRC-90-0103, in response to Generic Letter 88-01 and NUREG 0313 Rev. 2. Detroit Edison has committed to the inservice inspection requirements for austenitic stainless steel welds in accordance with the guidelines of Generic Letter 88-01. All applicable welds have been classified according to NUREG 0313 Rev. 2 requirements with the required percentages of welds being included in this program. The applicable category (GL-88-01) is identified in the remarks column. All welds identified as augmented selections will only be examined by volumetric techniques (i.e. ultrasonic). All inspections will be performed utilizing procedures and personnel qualified to current Utility PDI Guidelines. Sample expansion, if required, shall follow the NRC Staff recommendations provided in NUREG 0313 Rev 2 (Relief Request RR-A29). Methods and criteria for crack evaluation and j
repair shall be in conformance with IWB-3600 of Section XI of the 1989 Edition of ASME Boiler and Pressure Vessel Code. The previous requirement (now superseded) i was an 80 month augmented inspection cycle per NUREG 0313 Rev. I with inspections being performed per IE Bulletin 83-02. Detroit Edison requested that Non-Safety
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ISI-NDE Program i
Rev.2 Change 0 Table A Page 50 Related, Category D welds be removed from GL-88-01 scope per NRC-92-090. The NRC response (TAC No. M84117,12-18-1992) modified the inspection interval such that inspection of the subject piping welds on a sampling basis of at least 10 percent of 1
the weld population be performed during each refueling outage.
NOTE 3 Per the EF-2 UFS AR Subsection 4.5.1.2.7, Detroit Edison had agreed to ultrasonically inspect the RPV Jet Pump liold Down Beams at each Reactor Refueling Outage until sufficient experience was gained to change the frequency ofinspection. If a cracked beam was detected, it would be replaced prior to return to power operation. Due to the failure of ajet pump hold down beam at another plant, SIL No. 330, Supplements 1 and 2, and RICSIL No. 065 were issued. As a result, during RF04 all jet pump hold down beams were replaced with beam assemblies that are less susceptible to IGSCC than the original assemblies. Subsequent UT and alternative inspections will be performed at future refueling outages based on industry experiences and the recommendations provided in IE Bulletin 80-07 and NUREG/CR-3052.
NOTE 4 External surfaces - 25% nozzles among each group of penetrations of comparable size I
and function.
NOTE 5 Component supports and the associated integrally welded attachments are selected for examination in accordance with Code Cases N-491-1 (Alternative Requirements for Selection and Examination of Component Supports) and N-509 Alternative Rules for the Selection and Examination ofIntegrally Welded Attachments).
NOTE 6
. Visual examination of snubbers covers only the snuboer unit, except for those snubber supports selected in accordance with Code Case N-491-1. The balance of the support (Integral and nonintegral attachments including lugs, bolting, pins, clamps, and suppon steel) will be visually examined in accordance with subsection IWF requirements.
NOTE 7 Per SIL 420 an inspection will be performed on the jet pump sensing lines and suppon brackets when convenient. This inspection will determine if the weld between the support brackets and the vertical run of the sensing line is intact. Additionally the inspection should concentrate on the jet pumps closest to the recirculation outlet nozzles.
ISI NDE Program Rev.2 Change 0 Table A Page 51 NOTE 8 Per NRC Information Notice No. 90-30 all dissimilar metal welds containing Inconel 600 series base materials, Alloy 82 and 182 weld butter, and/or filler metal shall be examined following the guidelines of SIL 455, Revision 1 Supplement 1 (effective 6-90). It is essential and required that all examinations be performed by the use of 45 and 60 refracted longitudinal waves for crack detection and sizing in the Alloy 182 material and the low alloy material. All scanning of welds will be performed in both an axial and circumferential direction followed by a 45 shear wave ifindications are identified using refracted longitudinal techniques. Examination of nozzle welds shall be extended into the area of Alloy 182 Weld Material ~ Buttering. The purpose of this additional / supplemental examination is to assure that Alloy 182 Butter Cracking in the nozzle bore has not occurred and extended into the low alloy nozzle material.
NOTE 9 Per SIL 433, Supplement 1, an Ultrasonic (UT) inspection of the entire shroud head bolt length was performed on the 48 shroud head bolts for evidence of cracking during RF04.
l Based on industry experience additional inspections will be performed at subsequent refuel outages.
NOTE 10 During RF-06 the Reactor Recirculation Pumps were modified to the 4th generation design configuration. This configuration was designed to mitigate known causes to shaft and cover cracking and provides for ultrasonic inspection of the shaft without requiring complete pump disassembly and removal. This change out also included change out of l
the rotating element to a welded impeller and added rotating baffle. In addition, the hydrostatic bearing was modified to a non-welded design. The need to completely disassemble is reduced by modification to the 4th generation configuration. The following augmented inspections will be performed if the pump is disassembled. Per SIL 415, a supplemental liquid penetrant or volumetric in:,pection of the suction splitters will be performed if visual inspections identify cracking of the suction splitters or attachment welds. Per RICSIL 038 and NRC Information Notice 89-20 inspections will be performed on the hydrostatic bearing and baffle plate. Inspection of the heater / cooler assembly should be performed if the pump is disassembled. Disassembly of the pump for inspections will be evaluated prior to each refuel outage based upon industry experience and hours of operation NOTE 11 Per SIL 474 a visual inspection will be performed on the steam dryer drain channel welds during refueling outages. The steam dryer assembly, dryer banks, and welds will be visually inspected each refueling outage.
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i ISI-NDE Program Rev.2 Change 0 Table A Page 52
)
NOTE 12 Per IE Bulletin 80-13, and SIL 289, Revision 1, Supplement 2, a visual inspection will be performed on the core spray internal piping each Refuel Outage. Inspection points will include those specifically identified in IE Bulletin 80-13 and SIL 289, Revision 1, Supplement 2. The inspection plan will follow the inspection recommendations provided
-in BWRVIP-18.
NOTE 13 Per SIL 462 inspection of the shroud support access hole cover will be performed at the end of the first 10-year interval or during the plants tenth year of operation. Subsequent reinspections will be based on industry experience.
NOTE 14 All Inservice Examinations of the Reactor Pressure Vessel Welds will be performed using both manual and mechanical examination techniques and will most likely be performed from the outside of the vessel. All examinations will be conducted in accordance with the requirements of Regulatory Guide 1.150, Revision 1, to the extent practical (Ref. NRC-87-0078).
Limitations' encountered which affect the examination volume as prescribed by ASME Section XI will be documented in an examination report as required by Regulatory Guide 1.150, Revision 1, Appendix A. Regulatory Guide 1.150, Revision 1, Appendix A, I
recommends the use of the 2 percent notch which penetrates the internal (clad) surface of calibration blocks for detection of near surface flaws in that region. This is the calibration and examination method that shall be used.
Indications, regardless of amplitude, will be recorded on tape during the mechanized examination for analysis. Similarly, signal responses will be scrutinized during the manual examination process and indications will be recorded for further analysis and resolution.
NOTE 15 Visual inspections for leakage required by ASME section XI Code Categories B-P, C-H, and D-B are performed using site procedures. Test Packages for all tests performed are developed utilizing the Inservice Inspection Classification Boundary Drawings listed on Table A-5-5.1 as the basis.
All components on the following systems are included in the Class 1 inspections: B21, B31, C41, El1, E21, E41, E51, G33, N21, P34.
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ISI NDE Program Rev.2 Change 0 Table A Page 53 All components on the following systems are included in the Class 2 inspections: Cl1, C41, E l 1, E21, E41, G41, G51, N11, N30, P34, T4804, T50.
All components on the following systems are included in the Class 3 inspections: El 1, P42,P44,P45,R30.
NOTE 16 Per RICSIL No. 059 and SIL No. 554 inspection of the top guide beams should be performed at grid locations where fuel and blade guides have been removed for other reasons. Inspection of selected grid locations will be performed during refueling outages.
Additionally, ultrasonic inspection should be considered if cracking is found or as recommended by SIL No. 554.
NOTE 17 Per SIL No. 551, inspection should be performed of at least 50% of the Jet Pump Riser Brace welds at each Refueling Outage. Per SIL No. 574 a visual inspection of the jet pump adjusting screw tack welds should be performed during each refueling outage.
Based upon acceptable inspection results, future inspections of at least 50% of the tack welds each outage is sufficient. Repairs if required will be performed in accordance with the recommendations of SIL No. 574 as appropriate. In addition, verification of contact will be performed on the restrainer screws and wedge assembly to the inlet mixer per the recommendations of RICSIL-078. These inspections will be performed in conjunction with the inspection of thosejet pump assemblies selected for examination. The extent of inspection frequency will follow the recommendation provided in BWRVIP-41.
NOTE 18 i
Per recommendation of SIL No. 571 augmented inspection of this stainless steel nozzles should be performed after 15 years of operation. The inspection boundary for this weld shall be extended to include all stainless steel material accessible for ultrasonic examination. Iflinear surface indications are found, ultrasonic examination should be used to determine crack depth.
NOTE 19 Visual inspection of the core shroud and shroud welds will be performed in accordance with the recommendations contained in BWRVIP "BWR Core Shroud Inspection and Flaw Evaluation Guideline"(BWRVIP-01) utilizing techniques detailed in BWRVIP
" Reactor Pressure Vessel and Internals Examination Guidelines"(BWRVIP-03). SIL No.
572, Rev 1 inspection recommendations have been superceded. Fermi 2 has committed to perform future inspections per the guidance of the BWRVIP. Visual inspections will be performed as an enhanced VT-1 inspection with the capability to resolve a 1/2-mil wire on the inspection surface. The BWRVIP has imposed additional guidelines for 1
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1 ISI-NDE Program Rev.2 j
Change 0 Table A Page 54 inspection based on years of operation, materials, and conductivity. Based on the above, during RF-06 a baseline inspection of the shroud welds (11-3,11-4, li-5, and 11-7) was completed (approximately 90% volumetric coverage) utilizing an augmented ultrasonic phased array technique with no indication of service induced flaws. Future Core Shroud inspections will be performed in accordance with the BWRVIP " Guidelines for Reinspection of BWR Core Shrouds (BWRVIP - 07). Evaluation of anomalies shall be per the BWR Core Shroud Evaluation Reports (BWRVIP-01 and GENE-523-A53-0494).
Additional references include SIL 572, Rev 1 RICSIL 054, Rev 1, RICSIL 068, RICSIL 077, Information Notices93-079 and 94-042 and Generic Letter GL 94-03. GL 94-03 required advanced notification to the NRC of the proposed plan for Core Shroud inspection, evaluation and/or repair.
l NOTE 20 Additional augmented examinations were performed during RF04 and changes were
{
made to inspection schedule for selected nozzle welds following Turbine Generator i
Event and subsequent RPV chemistry transient for detection of IGSCC initiation.
Note 21 The new containment inspection requirements of ASME Section XI 1992 in effect for the Second Ten-year inspection interval changed the way containment system piping (between the isolation valves) are classified for ISI. IWE-1220(d) specifies that containment system piping is exempt from IWE requirements but shall be examined in accordance with the appropriate classification specified in the construction Design Specifica ions. This varies from the assumptions made during the first interval, when no IWE requirements were imposed. Relief Request RR-A26 documents Detroit Edison's proposed alternative examination requirements.
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ISI-NDE Program Rev.2 ChangeO Part B Page 1 i
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3 i
PART-B INSERVICE INSPECTION-NONDESTRUCTIVE TESTING (ISI NDE) PROGRAM (PLAN)
FOR COMPONENT SUPPORTS I
1
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i
,'h ISI-NDE Program Rev.2 Change 0 Part B Page 2 j
1.0 Applicable Code The Fermi 2 Inservice Inspection Program (Plan) for Component Supports is implemented in accordance with the requirements of ASME Section XI of the Boiler and Pressure Vessel Code,1989 Edition with no Addenda.
2.0 Program Description Visual examination and functional testing requirements in accordance with ASME Section XI, Subsection IWF and Code Case N-491-1 for all component supports, except snubbers, are addressed in this section. The Inservice Inspection and Testing Program (Plan) for snubbers can be found in Part C of this document.
l All component supports to be examined will receive a VT-3 visual examination to determine the general mechanical and structural condition of the component support. Component supports with moving parts (other than pin connections) will also receive a VT-3 visual examination to determine conditions relating to the operability of the component support.
2.1 Supports To Be Examined In accordance with ASME Section XI, those component supports selected for examination k
shall be the supports of those components that are required to be examined under IWB, IWC and IWD. Components included in this program have been selected using the selection l
criteria specified in Code Case N-491-1 (reference ISI Evaluation 99-056).
1 Component supports determined not to have integrity for intended service are defined as inoperable. Additional examinations will be performed as def med per Code Case N-491-1
)
paragraph 2430 for component supports determined to be inoperable.
l Suppo:ts scheduled to be examined are listed in the Tables found in Section 5.0, Part B of this document.
1 r
i 3.0 Exemptions fI 1
Per ASME Section XI Code Case N-491-1 component supports exempt from the examination requirements of paragraph 2000 are those connected to components and items exempted from examination under IWB-1220, IWC-1220, IWD-1220, and IWE-1220. In addition, portions of supports that are inaccessible by being encased in concrete, buried underground or encapsulated by guard pipe are also exempt from the examination requirements of paragraph l
2000.
The Inservice Inspection Classification Boundary Drawings listed in Part A, Section 5.0 provide coded system functional drawings identifying all piping and components exempt from examination.
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ISI NDE Program Rev.2 Change 0 1
Part B Page 3 4.0 Relief Requests 1
No Relief Requests are included in Part B of this program for the second ten-year interval. If requests for relief become necessary they will be processed as described in Part A,4.0.
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5.0 inservice Inspection Program (Plan) Table B (Component Supports) 5.1 The accompanying table lists the component suppons to be examined during the second inspection interval. The table is divided into ISI Class - 1,2, and 3. The table contains the following information:
Code Class: The ASME Section XI Classification as determined in accordance with the Code of Federal Regulations (10CFR50.55a) and the guidance provided in Regulatory Guide 1.26, and NUREG 0800.
Unique Identification: Identifies the specific component support subject to examination. The identification number consists of the system number, isometric line number, and the support number.
Exam Method Selected: Identifies the code required method of examination (i.e.
visual) and the specific examination selected for each component shown (i.e. VT-3).
Type: Identifies the type of component support to be examined.
Relief Request: If applicable, indicates the request for relief applicable in accordance with 10CFR50.55a(g)(5)(iii). See Part-A, Section 4.0 of this document.
Interval: Refers to the 120 month inspection interval as discussed in Section 2.0 of this document.
Period: Defined as the 3 year period within the 120 month (10 year) interval when the specific examination is scheduled. There are 3 periods in each 10 year interval and they can vary by +/- 1 year collectively over the 10 year period. Each period contains the specific outage code indicating that the required exam has either been scheduled or completed.
Remarks: Is reserved for additional information to explain, amplify, or provide added details necessary to clarify the examination requirements.
5.1 List of Abbreviations: For def'mitions of abbreviations used in the following tables, refer to Part-A, Section 6.2 of this document.
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C
1 PART - C INSERVICE INSPECTION NONDESTRUCTIVE FXAMINATION (ISI-NDE) PROGRAM (PLAN)
FOR l
SNUBBERS i
ISI-NDE Program Rev.2 Change O Part C Page 2 1.0 Applicable Code The Inservice Inspection Program (Plan) for Nondestructive Examination (NDE) of snubbers will be implemented in accordance with the requirements of the EF-2 Technical Specifications and Section XI of the ASME Boiler and Pressure Vessel Code,1989 Edition. Upon completion of the improved technical specification project, inspection and test requirments will be transferred to the Technical Requirements Manual (TRM). Section 5.1 titled " Augmented Inservice Inspection Program for Snubbers TRM 5.1 will continue the use of the original Technical Specification requirements for testing of snubbers that have been previously approved for use at Fermi.
Justification for use of this alternative is provided in Relief Request RR-C3.
2.0 Program Description 2.1 Visual Examinations 2.1.1 Visual Examination Freauency/ Schedule.
Visual examinations are required on all safety related snubbers. These safety related snubbers are listed in Table C, provided in Part C, Section 5.0. Visual examination frequency is established per TRM 5.1,in accordance with the rules of Generic Letter 90-09.
9 Snubbers which appear inoperable as a result of the visual inspection shall be classified as Unacceptable and may be reclassified as Acceptable for the purposes of establishing the next visual inspection interval, provided that:
- 1) The cause for the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and
- 2) The affected snubber is functionally tested in the as-found condition and determined operable per the functional testing acceptance criteria.
The next visual inspection interval shall be determined based upon the total number of unacceptable snubbers found during the previous visual inspection, relative to the total snubber population.
ISI-NDE Program e
Rev.2 Change O Part C Page 3 2.1.2 Extent of Visual Examination All inspections are performed by personnel qualified and cenified in accordance with IWA-2300. The frequency of examination and testing of snubbers as described in the relief request applies only to the snubber unit. The balance of the component support (integral and nonintegral attachments, including lugs, bolting, pins, clamps, and support steel) is examined in accordance with the requirements ofIWF and Code Case N-491-1. This is due to the examination frequency of the snubber being more frequent than for component supports. The component sup,m podion of the snubber (lugs, bolting, pins, clamps and support steel) is examined separately once each interval per the required sampling rates of Code Case N-491-1.
2.1.3 Additional Examinations:
For the snubber support (integral and nonintegral attachments, including lugs, bolting, pins, clamps, and support steel) IWF-2430 provides the requirements for additional examinations to be performed when corrective measures are required (per IWF-3000).
d i
2.2 Functional Testing 2.2.1 Functional Testine Extent. Freauency. and Sampline Plan Functional testing of all safety related snubbers is scheduled to coincide with scheduled refueling outages at intervals of approximately 18 months. Testing is performed utilizing the sampling plans and acceptance criteria provided in TRM 5.1 The representative sample is randomly chosen from the various types of snubbers installed in the plant and is reviewed before beginning the testing to ensure, as far as practical, that the samples are representative of the various configurations, operating environments,
functionally tested is dependent upon the sample plan chosen and the number of failures.
k el l
If a snubber fails the inservice functional test requirements, that snubber, or its
- j replacement, shall be retested at the time of the next scheduled functional testing but shall not be included in the sample plan.
y,
ISI-NDE Program Rev.2 Change O Part C Page 4 2.2.2 Functional Testine AcceDiance criteria The functional testing acceptance criteria for all sizes of snubber includes verification that, activation (restraining action)is achieved within the specified range in both tension and compression; snubber bleed or release rate where required, is present in both tension and compression, within the specified range; and for mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel.
2.2.3 Functional Testinn Failure Analysis Snubbers which fail the inservice functional test require an engineering evaluatica and additional testing. Corrective measures are in accordance the requirements of TRM 5.1.
An engineering evaluation is performed on the components to which the inoperable e
i snubbers are attached. The purpose of this evaluation is to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers, in order to ensure that the components remain capable of meeting their designed service. An engineering evaluation shall be made of each failure to meet the functional testing acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the operability of other snubbers, irrespective of type, which may be subject to the same failure mode. If any snubber selected for functional testing either fails to activate or fails to move, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be functionally tested.
If additional samples are selected in accordance with the sampling plans, the selection of
(
snubbers shall be based upon the engineering evaluation.
1 v
2.3 Snuhber Service Life Monitorine The service life of various components are established by engineering information and shall be adjusted based upon test results and failure histories. The purpose of this program is to ensure that the service life of critical components is not exceeded during a period when the snubber is required to be operable.
?.
3.0 Exemptions - None.
H T
ISI-NDE Program Rev.2 Change 0
?
Part C Page 5 L
4.0 Relief Requests 4.1 Relief Request Description (Format)
Relief Requests are included where specific requirements of ASME Section XI are determined to be impractical. All Relief Requests include the following information:
4.1.1 Component Function /
Description:
Identification of the component (s) for which reliefis requested.
4.1.2 System
The applicable plant system (s) associated with the Relief Request.
4.1.3 ASME Code Class: The applicable ASME code classification 4.1.4 ASME Section XI Requirements: Identification of the specific ASME Section XI requirement that has been determined to be impractical.
4.1.5 Basis for Relief: Infonnation to support Detroit Edison's determination that the i
ASME Code requirement is impractical. The following data will be provided, if applicable:
Reference to the regulatory basis paragraph Detailed technical information supporting proposed alternate scope of e
examination.
Description of the proposed alternative examination's impact on plant safety f
(if any) and justification of any change in the overall level of plant safety.
4.1.6 Alternate Examination: Alternate examination (s) that are proposed will be identified. Both alternate examination (s) that are performed in lieu of the Section XI examination requirement (s) and alternate examination (s) that supplement partially completed Section XI examination requirement (s) will be identified.
4.1.7 Applicable Time Periml: A statement identifying when the Relief Request would apply during the inspection period or interval 0
1
't l
ISI-NDE Program i
Rev.2 Change O Part C Page 6 l
4.2 Relief Requests The following Relief Request (s) are included in this section:
RELIEF REQUEST NUMBER GENERAL DESCRIPTION RR-Cl DELETED RR-C2 DELETED RR-C3 Snubber Examination and Testing RR-C4 Rotation of Snubbers per Code Case N-508-1 l
t l
l l
l l
?
ISI-NDE Program Rev.'2 Change 0 Part C Page 7 SECOND INTERVAL RELIEF REQUEST RR-C3 COMPONENT FUNCTION / DESCRIPTION:
Snubber Examination and Testing Program SYSTEM:
All Systems included in the ISI NDE Program Plan ASME CODE CLASS:
Class 1,2, and 3 and MC ASME SECTION XI REOUIREMENTS:
ASME Section XI 1989, IWF-5300 specifies that snubber examination and testing be performed e
in accordance with the 1988 Addenda to ASME/ ANSI OM-1987, Part 4, using the VT-3 visual examination method described in IWA-2213.
HASIS FOR RELIEF:
Pursuant to 10 CFR 50.55a(a)(3)(i), Detroit Edison is requesting relief from ASME Section XI requirements to perform snubber examination and testing in accordance with 1988 addenda to Part 4 of ASME/ ANSI OM-1987 (OM-4). The requirements of OM-4 pre-date Generic letter 90-09, titled " Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions."
Fermi 2 had imposed examination and testing requirements in accordance with Technical b
Specification Surveillance Requirement 4.7.5 for all safety related snubbers including ASME Class 1,2,3 and MC. Functional testing provides a 95 percent confidence level that at least 90 0
percent of the snubbers operate within the specified acceptance limits. The performance of visual examinations is a separate process that complements the functional testing program and provides additional confidence in snubber operability. Visual examination requirements are based on NRC Generic letter 90-09, " Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions".
With implementation of the Fermi 2 Improved Technical Specifications, the snubber surveillance requirements of are being moved in their entirety to the Fermi Technical Requirements Manual (TRM 5.1 titled " Augmented Inservice Inspection Program for Snubbers").
Implementation of TRM 5.1 requirements (formerly TS 4.7.5) for snubber functional testing and visual examination has maintained a reliable snubber population. The TRM requirements 1
provide an equivalent level of quality and safety. These alternative requirements were previously
[
reviewed and approved by the staffin the Fermi Technical Specifications.
ISI-NDE Program Rev.2 i
Change O Part C Page 8 i
SECOND INTERVAL RELIEF REQUEST l
RR-C3 l
(continued)
ALTERNATIVES:
Detroit Edison proposes to continue to utilize the requirements of the Fermi Technical Specification 4.7.5 (as moved in their entirety to the Technical Requirements Manual) for visual examination and functional testing of all snubbers associated with ASME Class 1,2,3, and MC component supports. These alternate requirements provide an acceptable level of quality and safety.
l APPI.lCABLE TIME PERIOD Relief is requested for the second 10-year inspection interval.
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1 ISI-NDE Program Rev.2 Change O Part C Page 9 SECOND INTERVAL RELIEF REQUEST RR-C4 I
COMPONENT FUNCTION / DESCRIPTION:
All safety related snubbers SYSTEM:
All systems included in the ISI NDE Program ASME CODE CLASS:
Class 1,2, and 3 and MC t
ASME SECTION XI REOUIREMENTS: ASME Section XI,1989 Edition,IWA-7130 T
requires a documented program for replacement of all items within the scope of this division.
Footnote 1 indicates that reasons for replacement may include:
Discrepancies detected during inservice inspection a.
- b. Regulatory requirements change
- c. Design changes to imporve equipment service
- d. Design changes to improve reliability e.
Damage f.
Failure during service
- g. Personnel exposure j
- h. Economics
- i. End of service life j.
Discrepancies detected during maintenance BASIS FOR RELEF:
Pursuant to 10 CFR 50.55a(a)(3)(i) Detroit Edison is requesting relief from ASME Section XI requirements to provide a documented replacement program when snubbers are rotated for testing purposes. Detroit Edison is proposing to implement the alternative of ASME Code Case N-508-1 (copy attached). This Code Case is ASME approved indicating Code Committee consensus that the alternative will provide an acceptable level of quality and safety. Detroit Edison agrees with the Code Committee, since the purpose for the rotation is related to Code requirements to remove snubbers for acceptance testing and not equipment failure or design changes.
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ISI-NDE Program Rey,2 Change 0 Part C Page 10 SECOND INTERVAL REL.IEF REQUEST RR-C4 (continued)
The Code Case places the following restrictions on equipment rotation:
Items being removed and installed shall be of the same design and construction; a.
- b. Items being removed shall have no evidence of failure at the time of removal; c.
Items being rotated shall be removed and installed only by mechanical means;
- d. Items being installed shall previously have been in service; Preservice inspections and pressure tests shall be performed as required by IWA-7000; e.
f.
The Owner shall maintain a method of tracking the items to ensure traceability of inservice and testing records;
- g. Use of an NIS-2 form is not required except as provided in (i) below.
- h. Testing of removed snubbers and pressure relief valves, including required sample expansions, shall be performed in accordance with the Owner's test program.
- i. Repair or replacement of removed items, when required, shall be performed in accordance with IWA-4000 or IWA-7000.
ALTERNATIVE EXAMINATION:
Code Case N-508-1 (attached) allows rotation of serviced snubbers in place of the one being removed for testing and subsequent servicing as necessary.
1 APPI.lCAHI,E TIME PERIOD:
Y Relief is requested for the second 10-year interval.
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4 16 l
ISI-NDE Program Rev.2 Change 0 Part C Page i1 CASE N-508-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: May 11,1994 See NumencelIndex for expreaon and any renMirmation dates.
Case N 508-1 (6) Items being removed shall have no evidence of Rotation of Serviced Snubbers and Pressure Relief failure at the time of removal; Valves for the Purpose of Testing (c) Items being rotated shall be removed and Section XI, Division 1 installed only by tnachanical means; (d) Items being installed shall previously have Inquiry: What alternative rules to those stated in been in service; IWA-4000 (IWA-7000 for Editions and Addenda (e) Preservice iW-and pressure tests shall prior to the 1991 Addenda) may be used when, for be performed as reqmred by IWA 4000 (IWA-7000 the purpose of testing, snubbers and press' re relief for Editions and Addenda prior to the 1991 Adden-valves are rotated from stock and installed a com-da);
ponents (including piping systems) within the Secuon (f) 'Ihe Owner shall maintain a method of tracking XI boundary?
the items to ensure traceability of inservice inspec-tion and testing records; Reply: It is the opinion of the Committee that, as (g) Use of an NIS-2 form is not required escept as an alternative to the provisions of IWA-4000 (IWA-provided in (i) below; 7000 for Editions and Addenda prior to the 1991 (h) Testing of removed snubbers and pressure re-Addenda) and for the purpose of testing, snubbers lief valves, including required sample expansions, and relief valves may be rotated from stock and in-shall be performed in accordance with the Owner's stalled on components (including piping systems) test program; within the Section XI boundary provided the follow-(l) Repair or replacement of removed items, when ing requirements are met:
required, shall be performed in socordance 'with (a) Items being removed and installed shall be of IWA-4000 (IWA-4000 or IWA-7000 for Editions and the same design and construction; Addenda prior to the 1991 Addenda).
i
ISI-NDE Program Rev.2 Change O Part C Page 12 INSERVICE INSPECTION PROGRAM (PLAN) TABLES (SNUBBERS) 5.0 The following tables list the snubbers that are to be examined during each interval.
Snubbers are grouped by the system they are associated with. The tables contain the following information:
Component Number: A unique identification number for each snubber component support.
Snubber Type: Mechanical or Hydraulic design.
Quantity / Size: Number of snubbers at one location and the size of each snubber.
Building / Floor: The building and floor level location of the snubber.
Elevation: The elevation of the snubber.
Inaccessible: The accessibility of the snubber during power operation. An X placed in this column indicates the snubber is inaccessible during power operation.
ALARA Concerns: An X placed in this column indicates a snubber is located in a radiation area greater than 60 millirem /hr, Difficult to Remove: An X placed in this column indicates the snubber is difficult to access or remove for functional testing, 5.1 List of Abbreviations:
f For definitions of abbreviations used in the following snubber tables, refer to Part A, Section 6.2 of this document.
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ISI NDE Program Revision 2 ChangeO Part D Page 1 PART - D ADOPTED ASME CODE CASES INSERVICE INSPECTION-NONDESTRUCTIVE EXAMINATION (ISI-NDE) PROGRAM (PLAN) 2 T
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ISI NDE Program Revision 2 Change 0 lg Part D Page 2 J
The Code Cases listed in the table below may be used in conjunction with the ISI NDE Program or Fermi procedures that specify requirements for examination, testing, and repair or replacement of Code Items. The listed relief requests are approved in Regulatroy Guide 1.147, Revision 12. Some of the listed Code Cases have been approved with restrictions and may only be used as noted.
t CODE CASE SUHJECT RESTRICTIONS
- 1. N-356 Certification Feriod for Level Ill NDE None Personnel
- 2. N389-1 Alternative Rules for repairs, None Replacements, or Modifications l
- 3. N-416-I Alternative Pressure Test Requirements Additional surface examinations l
for Welded Repairs or Installation of should shall be performed on the root Replacement Items by Welding, Class 1, pass layer of butt and socket welds of 2, and 3 the pressure retaining boundary of Class 3 components when the surface examination method is used in accordance with Section III
- 4. N-427 Use of Section XI Code Cases in None Inspection Plans
- 5. N-432 Repair Welding Using Automatic or None 1
Machine Gas Tungsten-Arc Welding 4
(GTAW) Temperbead Technique i
j
- 6. N-437 Use of Digital Readout and Digital None measurement Devices for Performing Pressure Tests
- 7. N-457 Qualification Specimen Notch Location None for Ultrasonic Examination of Bolts and f
Studs
- 8. N-458 Magnetic Particle of Coated Materials Thickness measurements and weld joint contour of the pipe / component must be known and used by the inspector who conducts the UT exam.
- 9. N-460 Alternative Examination Coverage for None Class 1 and Class 2 Welds
- 10. N-461 Alternative Rules for Piping Calibration None
..l Block Thickness i
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- 11. N-485-1 Eddy Cunent Examination of Coated None i
Ferritic Surfaces as an Alternative to i!
Surface Examinatin
- 12. N-489 Alternative Rules for Level III NDE None Qualification Examinations j
- 13. N-490-1 Alternative Vision Test Requirements None for Nondestructive Examiners
- 14. N-491-1 Alternative Rules for Examination of None Class 1,2,3, and MC Component Supports of Light-Water Cooled Power Plants
- 15. N-498-1 Alternative Rules for 10-Year System l
Hydrostatic Testing for " Class 1,2, and
[
3 systems I
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ISI NDE Program Revision 2 ChangeO Part D Page 3 CODE CASE SUBJECT RESTRICTIONS
- 16. N-503 Limited Certification of Nondestructive None Examination Personnel
- 17. N-504-1 Alternative Rules for Repair of Class 1.
None 2, and 3 Austenitic Stainless Steel Piping
- 18. N-509 Alternative Rules for the Selection and A minimum 10% sample ofintegrally Examination of Class 1,2, and 3 welded attachments for each item in Integrally Welded Attachments each code class per interval shall be examined.
- 19. N-517 Quality Assurance Program None Requirements for Owners
- 20. N-522 Pressure Testing of Containment The test shall be conducted at the Penetration Piping peak calculated containment pressure and the test procedure should permit the detection and location of through-wall leakage in containment isolation valves (CIVs) and pipe segments
'g between the CIVs.
- 21. N-524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Ji 5
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