Text

Rev 4 to ISI Plan for Bngs First Insp Interval
	ML20129C401
Person / Time
Site:	Braidwood
Issue date:	10/17/1996
From:	; COMMONWEALTH EDISON CO.
To:	;
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ML20129C394	List: ML20129C392; ML20129C401;
References
	NUDOCS 9610240016
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{{#Wiki_filter:- - = - - ~ Revision 4 i INSERVICE INSPECTION PLAN ] FOR BRAIDWOOD NUCLEAR GENERATING STATION UNITS 1 AND 2 l 1ST INSPECTION INTERVAL i l Commercial Service Date Unit 1: July 29,1988 l Commercial Service Date Unit 2: October 17,1988 Braidwood Nuclear Station R.R. #1 Box 84 Braceville, Illinois 60407 Comed P.O. Box 767 Chicago, Illinois 60690 Jeh2;gggggggggg33 O PDR

I I i Revsion 4 BRAIDWOOD NUCLEAR STATION, UNITS 1 AND 2, INSERVICE INSPECTION PLAN, 1ST INTERVAL REVISION

SUMMARY

SHEET Effective Page(s) Rev. l i 4 11 to lii 3 1-1 3 2-1 to 2-8 3 3-1 to 3-2 3 3-3 4 3-4 to 3-5 2 3-6 2 3-7 to 3-8 2 3-9 3 3-10 to 3-14 2 3-15 to 3-21 3 3-22 2 3-23 2 3-24 to 3-25 4 3-26 3 3-27 2 3-28 to 3-30 2 3-31 2 3-32 3 3-33 to 3-35 3 3-36 1 3-37 to 3-86 0 4-1 3 4-2 4 4-3 to 4-18 2 4-19 1 4-20 to 4-21 2 4-22 to 4-27 0 5-1 to 5-2 3 5-3 to 5-4 2 5-5 to 5-6 3 5-7 to 5-12 2 5-13 to 5-17 3 5-18 to 5-21 2 5-22 3 5-23 2 5-24 3 5-25 2 5-26 to 5-31 3 6-1 to 6-3 2 6-4 to 6-5 1 7-1 to 7-3 1 7-4 to 7-7 2 7-8 to 7-9 1 8-1 3 9-1 to 9-7 3 10-1 3 i

-~ - Revision 4 Table 3.0-1 Relief Request Summaries (sheet 2 of 2) NR-21 Alternate Hydrostatic Pressure Test Requirements for ASME Class 1, 2 and 3 repaired or replaced components. NR-22 Alternate rules for 10 year Hydrostatic Pressure Testing for Class 3 systems. NR-23 Alternate Examination of Nozzle to Vessel Welds - Residual Heat Removal Heat Exchangers. NR-24 Alternative rules for the Inservice Inspection of the Pressurizer Surge Nozzle to Shell Weld and Nozzle Innel Radius Section. NR-25 Alternative rules for the Inservice Inspection of the Pressurizer Seismic Lug Welds. NR-26 Alternative rules for the Inservice Inspection of the Inaccessible Welds on Welded Attachments NR-27 Alternative rules for the Inservice Inspection of Reactor Vessel Nozzle to Vessel Welds NR-28 Alternative rules for the Inservice Inspection of Reactor Vessel Nozzle Inner Radius Sections (IRS) NR-29 Alternative rules for the Inservice Inspection of Reactor Vessel Circumferential Shell Welds, Lower Head Circumferential Weld and Shell to Flange Weld. NR-30 Alternative rules to Table IWC-2500-1, Category C-H: Pressure Testing of Containment Penetration Piping with attached nonclassed piping. NR-31 Limited Volumetric Examination of Reator Vessel Circumferential Shell Weld. i 3-3

Revision 2 RELIEF REQUEST NR-7 1. SYSTEM: Reactor Coolant i 2. NUMBER OF ITEMS: 6 Weld # Line # Unable to be Examined Reason for limited Exam 1RC-16-2 1RY01C-4" 16% Elbow inner radius, reducer geometry 1RC-16-5 1RY018-6" 16% Elbow inner radii 1RC-32-1 1RYO3AA-6" 16% Elbow inner radius, nozzle geometry 1RC-32-7 1RYO3AB-6" 16% Elbow inner radius, nozzle geometry 1RC-32-13 1RYO3AC-6" 16% Elbow inner radius, nozzle geometry 1RC-35-1 1RYO2A-6" 16% Elbow inner radius, nozzle geometry 3. A.S.M.E. CODE CLASS: 1 4. A.S.M.E. SECTION XI REQUIREMENTS: Table IWB-2500-1, Examination Category B-J, Item B9.ll requires volumetric and surface examination of the areas described in Figure IWB-2500-8 for essentially 100% of the weld length. 5. BASIS FOR RELIEF: The above listed welds have interfering conditions on each side of the weld. These interferences can cause: poor coupling of the transducer, limited movement of the transducer, redirecting of the sound beam and, in some cases, complete restriction of a particular scan. These conditions sufficiently limit the axial scans so as to leave the listed percent of weld length uninspected. 6. ALTERNATE TEST METHOD: All welds shall receive the required Section XI surface examination in addition to the best effort ultrasonic examination. 7. JUSTIFICATION: The estimates of weld lengths unable to be examined are extremely conservative and are actually a percent of weld length for which there is less confidence that the entire required weld volume will be examined. Based on the required surface and leakage tests the structural integrity of this weld shall be assured. 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month interval. 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91. i 3-13

Revision 0 i J RELIEF REQUEST NR-20 (Cont.) 5. BASIS FOR RELIEF (Cont): This relief request will only apply to Braidwood Unit 1 steam generators that will be replaced in 1998. The IRS of the nozzles on the new steam generators that will be installed in Braidwood Unit 1 will be subject to the applicable requirements of ASME Section XI for the second ten-year Inspection Interval. 6. ALTERNATE TEST METHOD: Periodic visual examination (VT-2) of the nozzles' outside surface will be performed in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and Table IWC-2500-1, Examination Category C-l H, including applicable Code Case (s). i Prior to the Braidwood Unit 1 steam generators replacement, if future IRS i examinations of Braidwood Unit 2 and Byron Unit 2 steam generator nozzles (primary and feedwater nozzles) reveal flaws that exceed the applicable acceptance criteria of ASME Section XI IWB-3500, appropriate Braidwood Unit 1 steam generator nozzle IRS will be examined in accordance with the applicable requirements of ASME Section XI. 7. JUSTIFICATION: Periodic IRS examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-D and Table IWC-2500-1, Examination Category C-B on Braidwood Unit 2 and Byron Unit 2 steam generator nozzles (primary and feedwater nozzles) will serve to provide a reasonable sample from which to assess the integrity of the steam generator primary J nozzles and feedwater nozzles. Currently, there are no plans to replace the j steam generators in Braidwood Unit 2 and Byron Unit 2. 8. APPLICABLE TIME PERIOD: This relief request will be required for the first ten-year Inspection Interval. 9. REFERENCES,1 None. 10. APPROVAL STATUS: Relief granted per SER dated 9/1/95. 3 J d ) 3-38

.m 1 4 Revision 0 i RELIEF REOUEST NR-21 (Cont.). Little benefit is gained from the added challenge to the piping system provided by an elevated pressure hydrostatic test when compared to an operational test. The piping stress experienced by a hydrostatic test does not include the significant stresses associated with the thermal growth and dynamic loading during operation or 1 a design basis event. Therefore, the system is more likely to leak at operating conditions, due to operational dynamic and thermal loading, than during the careful, slow pressurization associated with a hydrostatic test. The acceptability of performing nominal operating pressure tests in lieu of j hydrostatic tests is also supported by the recent approval by the Board of Nuclear Codes and Standards of ASME Code Case N-416-1, " Alternate Pressure Test Requirement 4 for Welded Repairs or Installation of Replacement Items by Welding for Class 1, 2, l and 3 Systems, Section XI, Division 1". This Code Case allows a system leakage test at nominal operating pressure and temperature, in accordance with IWA-5000 of the 1992 Edition of Section XI, to be performed in lieu of a hydrostatic test, provided I that Non Destructive Examination (NDE) of the weld (s) is performed in accordance with the methods and acceptance criteria of the applicable subsection of the 1992 q. Edition of Section III. Based on the above, Braidwood Station requests relief from the ASME Section XI Class 1, 2, and 3 repair / replacement elevated pressure hydrostatic testing requirements. J T PEoPOSED ALTERNATE PROVISIONS As an alternate to the existing ASME Section XI requirements, Braidwood Station will adopt the provisions of Code Case N-416-1, as approved by the Board of Nuclear Codes and Standards, with additional NDE requirements. Listed below are the proposed alternate provisions to be performed, which is a summary of Code Case N416-1 requirements with additional NDE requirements imposed by Braidwood Station. 1 A VT-2 visual examination will be performed at nominal operating pressure and + temperature in conjunction with a system leakage test in accordance with IWA-5000 of the 1992 Edition of Section XI. The examination will be performed prior to or immediately upon return of the component to service. i Non-Destructive Examination will be performed on the repair / replacement welds + or welded areas with the methods and acceptance criteria of the applicable Subsection of the 1992 Edition of Section III. In addition, when NDE is required by ND-5222 for Class 3 components, an additional surface examination will be performed on the root (pass) layer. A surface examination will also be performed on Class 3 socket / fillet welds. The use of this relief request shall be documented on the applicable NIS-2 Form. If the previous version of Code Case N-416 were used to defer a Class 2 hydrostatic test, the deferred test may be eliminated when the requirements of this relief request are met. In addition, the NDE methods and acceptance criteria of the Code Edition and addenda used for the repair must be reconciled to those of the 1992 Edition of Section III. APl],ICABLE TIME f_ERIOD Relief is requested for the first ten-year inspection interval of the Inservice Inspection Program for Braidwood Unit 1 and Unit 2. APPROVAL STATUS: Relief granted per SER dated 9/1/95. 3-40

Revision 0 RELIEF REQUEST NR-22 COMPONENT IDENTIFICATION Code Classes: 3

References:

IWD-5210 IWD-5223 Examination Categories: D-A, D-B, D-C Item Numbers: Dl.10, D2.10, D3.10

== Description:== Alternate Rules for 10 Year Hydrostatic Pressure Testing for Class 3 Systems Component Numbers: All Class 3 Systems subject to Hydrostatic Testing per IWD-2500. CODE REQUIREMENTS IWD-2500 requires elevated pressure hydrostatic tests (VT-2) to be performed each 10 year inspection interval on ASME Class 3 pressure retaining components in accordarce with IWD-5223. BASIS FOR RELIEF Elevated pressure hydrostatic tests are difficult to perform and often represent a true hardship without benefits gained. Some of the difficulties associated with 10 year system hydrostatic testing include: Complicated or abnormal valve line-ups to provide system draining, filling, venting, and system isolation. Relief valves with setpoints lower than the hydrostatic test pressure must be blocked closed, removed and blank flanged. This process requires draining, refilling of the system prior to the test and draining, valve restoration, and refilling once more for system restoration Improper blocking or gaging can result in damage to the relief valve. Valves that are not normally used for isolation are often required to provide pressure isolation for a hydrostatic test. In order to provide tight isolat:en, time consuming valve maintenance would be required prior to a hydrostatic test. The radiation exposure required to perform hydrostatic testing is quite high in comparison to operational pressure testing due to time required for valve manipulation, filling and venting, valve maintenance, etc. At hydrostatic test pressures required by ASME Section XI, 10t and 25i over the piping design pressure, a hydrostatic test does not induce significantly more stresses in the system than in a system operational test. Also, the system stresses associated with the hydrostatic test do not compare to the stress associated with thermal growth and dynamic loading during design basis events. Therefore, littic benefit is gained from the hydrostatic test over the nominal operational pressure test. Industry experience, which Comed Stations experience supports, indicates that most through wall leakage is detected during system operation as opposed to hydrostatic j testing at elevated pressures. j l 1 3-41 1

. - =.. -. ~... _. -.. - i l Revision 0 RELIEF REQUEST NR-22 (Cont.) These arguments are also supported by ASME Code Case N-498-1, " Alternate Rules for f 10 Year Hydrostatic Pressure Testing for Class 1, 2 and 3 Systems, Section XI, 1 Division 1" and ASME Code Case N-498, " Alternate Rules for 10 Year Hydrostatic Pressure Testing for Class 1 and 2 Systems, Section XI, Division 1". Code Case N-498-1 has been reviewed and approved by the Board of Nuclear Codes and Standards (BNCS). Code Case N-498 for Class 1 and 2 systems had previously been approved and accepted for industry use in Regulatory Guide 1.147, Revision 10. Based on the above, Braidwood Station requests relief from the ASME Section XI Class 3 10 Year System Hydrostatic Pressure Testing requirements. PROPOSED ALTERNATE PROVISIONS A system pressure test with a VT-2 visual examination will be performed with the Class 3 system pressurized to a test pressure equal to nominal operating pressure. The visual examination will be conducted after the system has been pressurized to test pressure for a minimum of 10 minutes for noninsulated components or 4 hours for insulated components prior to examination. The system will be maintained at test pressure for the duration of the VT-2 visual examination. Hydrostatic test instrumentation requirements of IWA-5260 are not applicable as test parameter ) recording is performed by normal operating system instrumentation or equivalent. 1 The system pressure test will be conducted at or near the end of the inspection interval or during the same inspection period of each inspection interval. The boundary subject to test pressurization and VT-2 visual examination during the system pressure test shall extend to all Class 3 components included in those portions of systems required to operate or support the safety system function up to and including the first normally closed valve (including safety or relief valve) or valve capable of automatic closure when the safety function is required. APPLICABLE TIME PERIOD Relief is requested for the first ten-year inspection interval of the Inservice Inspection Program for Braidwood Unit 1 and Unit 2. APPROVAL STATUS: Relief granted per SER dated 9/1/95. l l l J 3-42

_. _, _ _ ~ _. _. _ _ _ _ -. _ _. _. _ _ _ _ _ _ _ _ _ I Revision 0 RELIEF REQUEST NR-23 CCHPONENT IDENTIFICATION Code Classification: 2

References:

IWC-2500-1 IWC-2420 IWC-2430 IWC-3000 l 1. NUREG/CR 4878, " Analysis of Experiments on Stainless Steel Flux Welds," April 1987. j .i 2. T. W. Simpkin (Comed) letter to Dr. Thomas E. Murley (10SNRC), "Braidwood Station. Unit 2 Flow Evaluation for RHR Heat Exchanger Nozzle to Shell Welds," dated November 13, 1991 I 3. Robert M. Pulsifer (USNRC) letter to Thomas J. Kovach (Comed), " Residual Heat Removal Heat Exchanger Nozzle to Shell Welds (TAC No. M82087)," ) dated November 21, 1991 4. Harold D. Pontious, Jr. (Comed) letter to USNRC Document Control Desk, " Supplemental Information Regarding the Fracture Mechanics Evaluation of .) Residual Heat Removal System Heat Exchanger Inlet and Outlet Nozzle to Shell Welds," dated November 9, 1994 5. Denise M. Saccomando (Comed) letter to USNRC Document Control Desk, " Supplement to Fracture Mechanics Evaluation of Residual Heat Removal l System Heat Exchanger Inlet and outlet Nozzle to Shell Welds," dated December 20, 1994 6. George F. Dick, Jr. (USNRC) letter to D. L. Farrar (Comed), " Residual Heat Exchanger Nozzle Welds, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (TAC Nos. M90894, M80895, M91408 and M90840), dated February 3, 1995

Attachment:

1. Residual Heat Removal ( RHR) Heat Exchanger Nozzle to L Vessel Detail i Examination Categories: C-B ' Item Numbers: C2.21

== Description:== Alternate Examination of Nozzle to Vessel Welds - Residual Heat Removal Heat Exchangers Component Numbers: Component Weld Nunbers 1RH02AA RHXN-01, RHXN-02 1RH02AB RHXN-01, RHXN-02 2 RH02AA RHXN-01, RHXN-02 2RH02AB RHXN-01, RHXN-02 CODE REQUIREMENT Subsection IWC, Table IWC-2500-1, Examination Category C-B, Item C2.21 requirer, volumetric and surface examination of the Nozzle to Shell welds of the regions i described in Figure IWC 2500-4(a) or (b), for nozzles without reinforcing plate in -vessels >H in. nominal thickness. Examinations shall be conducted on nozzles et j terminal ends of piping runs selected for examination under Examination Category C-F 3-43

Revision b RELIEF REQUEST NR-23 (Cont.) each inspection interval. In cases of multiple vessels of similar design, size, and service, the required examinations may be limited to one vessel or distributed among the vessels. Per IWC-2430, additional examinations are required when an examination detects indications exceeding the allowable standards of IWC-3000. These additional examinations shall be extended to include an additional number of similar components (or areas) within the same category. Per IWC-2420, if component examination results require evaluation of flaw indications in accordance with IWC-3000, and the component qualifies as conditionally acceptable for continued service, the areas containing such flaw indications shall be reexamined during the next inspection period listed in the schedule of Inspection Program B of IWC-2412. If the reexamination reveals that the flaw indications remain essentially unchanged for the next inspection period, the component examination achedule may revert to the original schedule of successive inspections. BASIS FOR RELIEF History: The RHR Heat Exchangers at Byron and Braidwood were manufactured by Joseph Oats Corporation in 1975 per the requirements of ASME Section III, 1974 Edition, Summer 1975 Addenda, Subarticle NC3200, Alternate Design Rules for Vessels. The nozzles and shell are fabricated from SA240 type 304 stainless steel material. The RHR heat exchangers tube side is Code Class 2 and the shell side is Class 3. The nozzle to shell welds were not required to be volumetrically examined during fabrication and only liquid penetrant examinations were performed on the final surfaces of the weld. During the preservice inspections of the Byron and Braidwood components, relief was requested from performing volumetric examinations of the nozzle to vessel welds due to inherent geometric constraints. The fillet weld located directly above the l nozzle-to-vessel weld is an obstruction to the proper movement of the inspection instrumentation transducer. These constraints limited the ability to perform a meaningful UT. These relief requests, NR-14 for Byron Unit 1, NR-13 for Byron Unit 2, 1NR-12 for Braidwood Unit 1, and 2NR-12 for Braidwood Unit 2 were approved by the NRC in Byron SSER 7, page 16 and Braidwood SSER 5, page 6-2. Relief Requests NR-12 for Byron and NR-12 for Braidwood were included with the First Ten Year Interval ISI Program Submittal. These relief requests sought the same Code inspection exemptions for the nozzle to shell welds as did the preservice relief requests. Relief for the nozzle to shell weld examination was denied and NRR requested a best effort UT of the nozzle-to-vessel welds be conducted. The initial UT inspection (1991) performed on the Braidwood Unit 2 "A" RHR heat exchanger found indications which exceeded the ASME Section XI 1983 Edition, Summer 1983 Addenda, Subarticle IWC-3000 allowable limits. The indications which exceeded the acceptance standards of IWC-3000 were subjected to further evaluation in accordance with ASME Section XI Subarticle IWB-3600. The required Fracture Mechanics Analysis was submitted to the NRC (Reference 2) and the indications were found to be acceptable for continued service (Reference 3). Additional examinations were performed in 1992 for Byron Unit 2 and Braidwood Unit 1 heat exchangers, in 1993 for Byron Unit 1 and Braidwood Unit 2 heat exchangers, in 1994 for Braidwood Unit 2 heat exchangers, and in 1995 for Byron Unit 2 heat exchangers. All i examinations confirmed the existence of fabrication flaws in the nozzle to vessel welds. The examination results from the inspections performed in 1994 at Braidwood Unit 2 included flaws on the outlet nozzle weld of the 2B RHR vessel which exceeded the 60-3-44

Revision 0 RELIEF REQUEST NR-23 (Cont.) acceptance criteria. The size change from previous inspections was attributed to enhancement in the volumetric examination technique. An ASME section XI repair by excavation was completed; the unacceptable flaws were removed. Safety significance: The RHR Heat Exchanger welds are within a class 2 system, on a moderate energy line which operates at a relatively low pressure ( 400 psig). This operating pressure is below the design pressure (600 psig) used for allowable flaw size calculations in the Fracture Mechanics Analysis. The actual induced piping loads on the nozzles are less than 60% of the design loads used by the allowable flaw size calculations. Observations made of the excavation areas on the Braidwood Unit 2 "B" Outlet RH HX Nozzle ( 2 RHX-01) repair verified that the indications found in the RH HX Nozzles are fabrication flaws, slag, incomplete fusion and excess porosity. No service induced flaws were found. A hydrostatic test was performed by the manufacturer, after fabrication, for all vessels at a pressure of 803 psig. Another hydrostatic test was performed in the field, after installation, at a pressure of 750 psig for Braidwood Unit 1 and 800 psig for Braidwood Unit 2 with no leakage noted from these regions. Pressure is the dominant load on the nozzle weld. The hydrotests have demonstrated that these nozzle welds can withstand almost double the operating pressure, without structural failure, despite the presence of the fabrication flaws in the weld. The Fracture Mechanics Analysis shows that these nozzles have a large flaw tolerance because of material ductility, flexibility (thln walled), and the reinforcement j provided by the fillet weld. It has also been shown (Reference 1) that the fracture toughness of flux welds is higher than that used in the allowable flaw size calculation performed as part of the fracture mechanics evaluation. A finite element analysis was performed and submitted to the NRC for review (Reference 4). The analysis was subsequently supplemented (Reference 5). The results of this analysis show that the inside diameter of the nozzle is in compression and the outside diameter (O.D.) is in tension. Consequently, any service induced flaw wculd b, expected to initiate at the O.D. of the nozzle where the weld membrane stresses are in tension. All the fabrication flaws exist within the areas shown in the analysis to be in compressive or negligible stress and are not subject to propagation. The NRC review of the finite element analysis is documented in Reference 6. The objective of the Inservice Inspection Program is to find " service induced flaws" before they become safety significant. A service induced flaw will initiate as a surface flaw at the nozzle O.D., as discussed above, so a Penetrant Test will be more likely to detect a service induced flaw than a volumetric exam.

Also, due to the low stresses present and given the fracture toughness of stainless steel, leakage from the joint would likely be detected before a leak would occur.

A VT-2 examination is being conducted on all RHR Heat Exchangers once per inspection period as required by ASME Section XI code Item C2.33. Perforraance of surface examinations each inspection period will provide the best means for detection of service induced flaws and provide assurance that a service induced defect will be identified prior to component failure. Ultrasonic examinations for the RHR nozzle-to-vessel weld will not provide detection capabilities of service induced flaws beyond that provided by surface examination. Additionally, performance of the ultrasonic examinations will require extensive labor resources, unnecessary radiation exposure to the examiners and add significant costs to Commonwealth Edison without a commensurate increase in quality or public safety. 3-45

-- -.- ~ _ Revision 0 RELIEF REQUEST NR-23 (Cont.) Justification: The volumetric examinations and the repair completed at Braidwood Unit 2 characterize these flaws as fabrication defects, and not service induced cracks. Additionally, the Fracture Mechanics Analysis predicts negligible crack growth. The Fracture Mechanics Analysis also revealed that the inside nozzle surface is in i compression and the outside surface is in tension. Therefore, a Section XI surface examination is an adequate test to verify the structural integrity of the welds. PROPOSED ALTERNATIVE EXAMINATIONS i The nozzle-to-vessel welds on the A and B RHR Heat Exchangers for Braidwood Units 1 ( and 2 will receive a Section XI surface examination each inspection period. In addition, a visual examination (VT-2) shall be performed each inspection period on all pressure retaining components. APPLICABLE TIME PERIOD j i This relief will be required for the first 120 month inspection interval. APPROVAL STATUS Relief granted per SER dated 2/29/96. Relief granted provided that volumetric l examinations on a sample of RHR no.121e-to-vessel welds (one nozzle per unit) be performed during the next inspection interval. 1 i 1 [ l 4 3 i f 1 3-46

Revision 0 NT,-2 3 23" 1/4" x 45* Chamfer Not to Scale r % TYP. (4) Corners o l l' 1.25" Thick Plate 20"

l------r------

l l' T s / E Shell Axis DETAIL A 5/8" \\ 5/8" V'88*1 W^11 T = 0.875" h', 1 l n 'j J 'd Nozzle (Type) i) N = 0.375" ..se. b f E ' [f'-- 1 = ' ' ' ~ N ~ o ,\\ l* W 3/8" Reinforcement Pad ( l W = 1.25" 3/16" l (See Deatail A) RHR HEAT EXCHANGER NOZZLE CONFIGUI6:. TION 3 47

c ___-.~. .. _ ~, Revision 0 RELIEF REQUEST NR-24 COMPONENT IDENTIFICATION CODE CLASS (ES): (1) One REFERENCE (S): Subsection IWB, Table IWB-2500-1 EXAMINATION CATEGORY (IES): B-D ITEM NUMBER (S): B3.110 & B3.120 - DESCRIPTION: Pressurizer Surge Nozzle-to-Vessel Weld and Pressurizer Surge Nozzle Inner Radius Section COMPONENT NUMBER (S) : 1PZR-01-N1 l 2PZR-01-N1 4 ASME CODE SECTION XI REQUIREMENT 4 Perform voluma* ' (JT) examination of the nozzle-to-vessel weld and the nozzle inner radius s 1. BASIS FOR RELIEF In accordance with 10CFR50. 55a (g) (5) (iv), relief is requested on the basis that compliance with the Code requirements would result in hardship or unusual' difficulty without a compensating increase in the level of quality and safety. Braidwood Units 1 and 2 pressurizer nozzles are welded to the vessel heads j (Attachment 1). Each pressurizer has a single surge nozzle in the lower head. To j perform UT examinations on these areas, the outside surface of the lower vessel head, which is the optimal scanning surface, must be accessible. This optimal j scanning surface is made accessible by removing the lower press rizer head I insulation. The impact of removing the lower head insulation is discussed below. ) The lower head of the pressurizer is covered by 4 inches of multi-layered stainless steel mirror insulation. To remove the insulation, the 78 pressurizer heater cables I. would have to be disconnected (Attachment 2). In addition, each of the 78 convection stops, which are riveted to the insulation, would have to be cut so that the insulation could be removed (Attachment 3). i-Previous attempts to acquire this data at another Comed plant have proven unsuccessful. During previous outages, an attempt was made to modify the insulation on the lower head of the Byron Unit 2 pressurizer to allow inspection access without 4 full insulation and heater cable removal. The insulation group worked for three shifts per day for five days to remove this insulation. The groups used small grinders to cut the insulation from the nozzle to the first ring of immersion heaters. After this work was completed, the bottom head insulation was lowered until stopped by the heater connections. These actions did not result in sufficieat access to conduct the required examinations. Further actions to provide access were -determined to be impractical. The insulation was replaced and the cut areas were 3 J covered. j j Examination of the nozzle to vessel weld and the nozzle inner radius would result in j limited examination coverage. Even if the insulation were removed, full ultrasonic i examination coverage of the surge nozzle-to-vessel weld can not be achieved. The l pressurizer surge nozzle geometry limits transducer contact. 1 Consequently, scanning on the nozzle side of the weld is impracticable. The heater penetrations obstruct scanning from the shell side of the weld The estimated ] coverage would only be approximately 60% of the weld volume. Regarding '.he nozzle 3-48 ~..

i Revision 0 l Relief Request NR-24 (Cont.) l inner radius, only limited ultrasonic examination of the nozzle inside radius section would be achievable from the outside surface with the insulation removed. I The complex geometry of the " blend region" is not conducive to typical UT 1 examination techniques. A lindted examination would be possible if ultrasonic scanning were conducted from the nozzle. However, due to the complex geometry of t the nozzle, the resulting coverage would provide very limited data from which to assess the condition of the surge nozzle inside radius section. The limited data 3 obtained from these examinations does not provide a compensatory increase in quality l and safety to justify the hazards of personnel radiation exposure incurred to obtain the data. 1 The radiation exposure to plant personnel for insulation removal, surface l preparation, and inspection is estimated to be 154 person-rem. To provide a basis ] for the dose estimates, a survey was conducted during the Braidwood A2R05 outage on j March 16, 1996. This survey shows a 500mR contact dose rate on the lower head } insulation with a general area rate of over 200mR. The primary work of disconnecting the heater cables, removing insulation, surface preparation, and inspection would occur in an area approximately 1 foot from the surge nozzle. After. the insulation is removed, the rates shown in the survey would increase. Lead shielding would not be practicable because the shielding would have to be placed on the surfaces that require work. 3 - Estimated Dose for PZR Surge Nozzle and Nozzle Inner Radius Examination. I Activity Man Hour Dose Rate Accumulated Dose (R) j Estimates * (R/hr)2 j Scaffolding 98 0.150 14.7 l Cable Disconnection / Replacement 412 0.250 103 j Insulation Removal / Replacement 140 0.250 35 l Surface Preparation 1 0.250 0.25 Examination 4 0.250 1 f Total 655 153.95 i f ' Time estimates established by W.A. Pope Company, the primary contractor, and Raytheon Engineers & Constructors. the inspection organization. l Whole body dose rate estimates based on location of worker's trunk for specified j work in required area at about 1 foot from surge line. 1 PROPOSED ALTERNATIVE EXAMINATION i i 1 The option of examining the pressurizer surge nozzle-to-head weld and nozzle inside j radius section from the inside surface has been addressed and determined to be impractical. The inside surface of the pressurizer surge nozzle is accessible only from the manway. Removal and reinstallation of the manway would incur significant i radiation exposure to plant personnel, which is estimated to be approximately 2 person-rem for Braidwood Unit 2. Braidwood Unit 1 would incur more dose to gain access to the pressurizer due to a diaphragm seal welded in the manway. Most importantly, baffle plates internal to the pressurizer would prohibit access to the debris screen and the surrounding inside surfaces of the nozzle for a meaningful VT-1 examination. To ensure compliance with 10CFR50.55a (g) (3), continued periodic visual examination (VT-2) of the nozzle inner radius area and nozzle to vessel weld will be performed according to the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P, including applicable Code Case (s). 1 i I 3-49 4

l Revision 0 Relief Request NR-24 (cont.) JUSTIFICATION Westinghouse Materials and Engineering Group has provided technical input to the basis for the exemption request for the nozzle to vessel weld and nozzle inner radius. This assessment discusses the structural integrity of the Braidwood Units 1 and 2 Pressurizer Surge Nozzle with respect to the nozzle to vessel weld and nozzle inner radius, and the need for the inservice inspection of these areas. The ass.essment includes three complimenting approaches which include inspection history, fracture assessment, and risk assessment. Each approach arrives at the same conclusion, which is that the inservice inspection of the nozzle areas do not significantly improve the confidence in the structural integrity of the pressurizer. Inspection History: The surge nozzle inner radius for each pressurizer is subjected to a surface examination both before and after the deposit of the stainless steel cladding. The inspection before cladding included 100% UT. The inspection after cladding was performed after the manufacturer hydrotest and included a radiographic examination for both the nozzle inner radii and nozzle to vessel weld for acceptance to ASME Section III. For preservice inspection, a UT was conducted of the nozzle to vessel welds with no indications in excess of allowables in ASME Section XI table IWB-3512-1. The nozzle inner radii did not have a preservice UT conducted due to the fact that no technique was available. Preservice relief request 1NR4 (Unit 1) and 2NR4 (Unit 2) were granted for the nozzle inner radius. For inservice inspection of the surge nozzles, access restrictions and the radiological concerns preclude contact examinations from the inside of the pressurizer. This leaves the only option to perform the examination from the nozzle outside surf ace blend region as described previously in the " Basis for Relief" A survey was conducted by the Westinghouse Owners Group, where it was discovered that roughly half of the plants surveyed have sought and received relief from volumetric examinations for the aforementioned reasons. Those that have been carrying out surge nozzle inspections have not reported any indications. Specifically, 21 inspections have been completed, 9 by using UT methods, with no reported indications. While this finding in itself is not sufficient to prove there is no need for further inspection in these areas, it is consistent with the other findings here, in that no concerns are evident with flaws in this region at the beginning of service, and there are no known mechanisms for cracks to initiate during service. Fracture Assessment: Westinghouse conducted fracture evaluations of the Braidwood surge nozzle inner radius and nozzle to vessel weld regions to determine the sensitivity of this region to the presence of a flaw. The full set of design transients was considered, and the most limiting event was found to be the heatup and cooldown, which can involve insurges of cooler water into the bottom of the pressurizer. The cooler water has a higher density than the water in the pressurizer before the insurge, and therefore mixing cannot be guaranteed. The worst case where no mixing occurs was addressed, and the maximum temperature difference between the loop and pressurizer of 320 F was assumed. Because the pressurizer is hot when the insurges occur, the fracture toughness value from the ASME Code Section XI Ku curve was found to be 200 ksiYin. The entire range of times during the insurge events was considered along with all the other design transients, and the stress intensity factor never exceeded the toughness, regardless of the size of the postulated crack. These results are summarized in Attachments 1 and 5. Therefore, the structural integrity of the pressurizer will not be affected by flaws in the surge nozzle inner radius or nozzle to vessel weld. 3-50

.. _ _. ~.. _ _ Revision 0 Relief Request NR-24 (Cont.) Risk Assessment: Westinghouse examined the effects of inservice examinations on the risk of failure due to cracking in the surge nozzle. From the fracture assessment it was determined that there is a very large tolerance for the presence of flaws in both the nozzle inner radii and the nozzle to vessel weld. Since the applied stress intensity factor does not exceed the fracture toughness, it could be argued that leakage would occur from a through wall flaw at the nozzle before any integrity problems would occur. There are no mechanisms of damage other than fatigue for the surge nozzle. Therefore, the only' scenarios of concern are for a flaw which was not found in the fabrication and preservice examinations to grow during service, or far a flaw to initiate during service and grow. The surge nozzle forgings for Braidwood Units 1 an's 2 were examined by both UT and MT prior to the cladding being applied. After cladding, the surge nozzles were required to be liquid penetrant tested to ensure the integrity of the cladding. The nozzle to vessel welds received both penetrant and volumetric (RT) during fabrication and UT during preservice examinations. With these examinations, it is extremely unlikely that a flaw of any size would be missed. Fatigue crack growth from any such flaw would be very small, and the fatigue assessments carried out to certify the design acceptance ensure that the fatigue loads during service are unlikely to initiate a flaw. Therefore the risk of failure-is very low, and is i unchanged whether or not inservice UT inspections are conducted. I conclusion: The assessments discussed above have shown that there is no compensating increase in quality or safety from ultrasonic inservice inspection of the surge nozzle and nozzle to vessel weld. Inspections which have been performed have not identified any indications at all in the entire population of Westinghouse plants, and the fracture assessment showed that the nozzle and nozzle to vessel weld have a very large tolerance for flaws. There are no mechanisms for the development of flaws during service, so that the risk of failure is not decreased by inservice inspection. A VT-2 inspection at pressure, along with Reactor Coolant System j Leakage Detection Systems ensure that through wall flaws would be identified prior to pressurizer structural integrity being compromised. APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval. 3 APPROVAL STATUS Relief granted per SER dated 5/3/96. However, if insulation is removed in the i future for any reason, the volumetric examinations will be required to be performed. 3-51

Revision 0 NR-24 1 Pressurizer / J t Debns screen t I (- i

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f:.'.. j 7 .'l ..s $'l

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2 li W [ @a __ I __ /' Head.. \\w! \\ a Heeter (Typ.) s) \\ ~ ~ w .i Blend region ~~ i i ,c Nozzle h .h ,N f,, insujetion gj ~ l s' I iI llll II l 1 Pressurizer Bottom Head Assembly 3-52

l Revision 0 NR-24 @e@@@@@ @ex @\\ g@@g@ g

g. @

tx"M @ g g e j eo@ = = <. g g e@ 7 h h Surge Nozzle h g@g@@@@O @ @g h h) Heate" Penetreua. ( t yp ) Immersion Heater Assembly 3-53

Revision 0 hm-24 M, VI Nr 8 g 1.1350 immersion Heater Pipe u up l l 1 4 i i I i l 1 - Outer Case s [ Washer -Pop-Rivet u I i i ~l l Sliding Con rection l Stop (78 Required) "_J__' Fie!d Instail Vl i i 's j ,l, d[iM --/ trom Vesse! Center g .3 s 's 's__[_ t Pivot ?om. t or deal r l / l l Bottom Head Insulation Details 3-54

. _ ~ d i i Revision 0 NR-24 i 4 4 t Stress intensity Factor Plot i 4 I' l 200 k 1._ i-am 1 \\ woo p - =1 4 2 ~ ._/ " 150 O y/_ / to . Z__ j tt,, .7 .b 4 l. sn ? 4, C 1M ._./. a .s l C - p 2.a__ 3 u) Sn sa r 4.. s.a / y _._. _p ~4 .._y. / . p_.,_ /. '.. 2 x, i s,. .e_ _2,_ Ap d ._. _.__p__._3 .._.p_ 4 _..._.q_ O O.1 02 03 0.4 0.5 0.6 0.7 0.0 0.9 M i i / Stress Intensity Factor Plot: Pressurizer Surge Nozzle Weld Inside Surface Flaw 3-55

I ) i 1 ) 1 Revision 0 NR-24 t i l l I I 1 l I l ) Stress intensity Factor Plot 200 xi-a= l u1 jnooHw::o(e-wrv) O y / ~ L tn /,.y _ Lt. _.1 __..a /_-- _ __...7 p.;__ --.. _ __... mC 13) .e_. . y.._2 s.. _ Twa

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Revision 0 RELIEF REQUEST NR-25 COMPONENT IDENTIFICATION Code Classes: 1

Reference:

IWB-2500-1 Examination Categories: B-H Item Numbers B8.20

== Description:== Alternate rules for the Inservice Inspection of the Pressurizer Seismic Lug Welds. Component Numbers: 1PZR-01-PSL-01, 1PZR-01-PSL-02 1PZR-01-PSL-03, 1PZR-01-PSL-04 2PZR-01-PSL-01, 2PZR-01-PSL-02 2PZR-01-PSL-03, 2PZR-01-PSL-04. CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-H, Item B8.20 requires surface or volumetric examination of Integrally Welded Attachnents to the Pressurizer figure IWB-2500-15. BASIS FOR RELIEF Relief is requested on the basis that compliance with the Code requirement would result in hardship or unusual difficulty without compensating increase in the level of quality and safety. Braidwood Units 1 and 2 Pressurizer _me.....Ac lugs are welded to the Pressurizer shell (reference Attachment 1). There are 4 seismic lugs per unit, located 90 degrees apart (reference Attac" ment 5). In order to perform examinations on the seismic lug welds, the outside surface of the lower vessel shell to lug area must be accessible. The exam surface is not accessible since it is covered by the seismic lug restraint j and lower Pressurizer shell insulation (reference Attachment 3 and 4). Also, the configuration of the Pressurizer coffin limits access to the seismic lugs. The impact of removing the seismic lug restraint, altering the Pressurizer coffin and removing the lower shell insulation is presented below. J The seismic restraint (Reference Attachment 1 and 2), which surrounds the lug, prohibits access needed to perform a meaningful surface exam. There are 4 restraints located about the 428' elevation, one for each lug, which were not designed for removal. The top of the concrete floor at this location is at 428' 3" elevation. This floor, which is 2'6" thick, interferes with access to 2 of the 4 lugs (Reference Attachment 2, 3 and 5). Also, the Pressurizer coffin itself severely limits access to remaining 2 seismic restraints (Reference Attachment 5). All of the restraints, which are embedded in the concrete, would require major modification to the existing Pressurizer coffin to allow for removal and access. This modification would require the redesign of the seismic restraint and Pressurizer coffin to allow for periodic removal and access to the seismic restraints. Implementation of this redesign would require significant engineering resources, construction resources and significant dose to plant personnel. i Only the upper panels were designed with clips to provide for removal. Insulation on the lower shell of the Pressurizer prohibits access needed to perform a meaningful surface examination of the seismic lug weld areas. The removal of the insulation covering the lower Pressurizer shell to seismic lug area will result in 3-57

~.-.. = I Revision 0 } RELIEF REQUEST NR-25 (cont.) i high radiation exposure to plant personnel. The insulation on the pressurizer i consists of panels which are fastened together. The lower panels are fastened J together with screws. To provide access from below would require scaffolding from the 401' elevation grating to the 428' elevation of the seismic restraint. Also, to remove the Pressurizer shell insulation would require removal of the screw J fasteners. Access to these screws is limited by the floor and Pressurizer i coffin (reference Attachment 3 and 5). As stated above, the insulation could be removed from the upper portions of the lugs. This can only be accomplished for 2 of I the 4 seismic lugs, because access is prohibited by the Pressurizer coffin j configuration (Reference Attachment 5)..The current configuration of the seismic I restraint also only allows limited access for visual examination. To provide j suitable access for all 4 seismic lug restraints would require major modifications j-and significant resources. l Even if the non removable insulation is removed (Reference Attachment 3, 4& 5), j full surface examination of the seismic lugs would not be achieved. The Pressurizer j coffin, concrete floor and seismic restraint geometry would greatly limit access to j all sides. The resulting coverage would only be a small percentage of the weld volume. The limited data obtained from these examinations do not provide a l compensatory increase in quality and safety to justify the hazards of personnel i radiation exposure to obtain the data. When the removable insulation panels are removed, it is estimated that 5.74" of surface per accessible lug will be i achievable. This accessible portion of surface can be visually inspected. It is i expected that only a best effort Liquid Penetrant (PT) exam can be performed on the j accessible exposed surfaces. Access and clearance interferences will limit how well 1 the surface of the examination volume can be prepped for the PT examination, i Because the examination is being performed on slightly rusted carbon steel j components, which will receive a best effort surface prep, that a white to pinkish ] back ground will be expected after developing. Even with a pinkish background, detection of relevant indications will still be possible. Also, bleed out.from the j lower edge of the non removable insulation will interfere with some of the accessible exam volume. This volume of interference will depend upon the amount of bleed out and will mask any relevant indication. i PROPOSED ALTERNATE PROVISIONS i l A VT-1 of the upper surfaces of the 2 accessible lugs when the removable insulation panels are removed. It is estimated that 5.74" per accessible lug (4" of the top j-and.87" of each side) will be achievable when the removable insulation panels are l removed. This is approximately 28.7% of the total exam volume for one lug. Also, a best effort surface inspection (Liquid Penetrant) will be performed on those portion i of the lug that are inspectable when the removable insulation panels are removed. In conjunction with the above proposed alternative technique, the periodic VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structural integrity of the Pressurizer shell. APPLICABLE TIME PERIOD This relief request will be required for the first ten-year Inspection Interval. APPROVAL STATUS Pending NRC review 3-58

l 1 Revision 0 NR-25 l (Drawing not to scale) d i l 3.5" m x / I E O O C M C Lug S. i \\ 4'0" Lug E.0" v 6 V 6.44" f 3[ 4.0"h 4 *",i'd@'c'"$ \\ B / \\ \\ 4.0' / /\\ O O O O O O 3 7g.. O O V ( / ( / h 1.25" A/s A 5.0- /s A ' l ' 4.0625" g ( 2.0" B i / l N Seismic Restraint Detail Pressurizer Seismic Lug and Restraint Detail 3-59

I Revision 0 NR-25 (Drawing not to scale) ) stils: PAcx f ~Y-j /\\ I I I 1. il l l I: 1 15"

s i

i i: V VIEW A-A I/ \\ N 17.0" r /\\ 3 o 15.0" I 1.5" l /\\ c : - + V V 4 . : WELDING SWDS. /\\ 2.3125" i12.D"/' 30.0" l; CONCRETE FIDOR h v l VIEW B-B Pressurizer Seismic Lug and Restraint Detail (continued) 3-60

Revision 0 NR-25 (Drawing not to scale) INSULAT [ON PANEL l REMOVAB 2 TYPE lf g Seismic Restraint 3.94" 1 i O omskt w Ik i E#i j___o 6.0" I h Sf[ f

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-;;.DONCRETE -l FLOOR e d 1 8 W* o$. c 5 E e ? ' V: Pressurizer Seismic Lug and Restraint Insulation Detail 3-61

Revision 0 (Drawing not to scale) l JConcriete1 Flo;or Pressurizer Seismic Restraint O O l O O O O 0 O + Shim Pack .\\, -r - %Y Yf lfLug l Insulation Panel Pressurizer Shell VIEW D-D Pressurizer Seismic Lug and Restraint Insulation Detail (continued) 3-62

l Revision 0 NR-25 (Drawing not to scale) j. PRESSURIZER COFFIN r '.g,~ R E s I Intccessible Lugs / ( Pressurizer l T y + Accessible Lugs T/ SLAB EL. 428'-3" Plan at Elevation 428'-7 1/2" UNIT 1 AS SHOWN UNIT 2 OPP. HAND 3-63

I

l l

l I . Revision 0 RELIEF REQUEST NR-26 COMPONENT IDENTIFICATION Code Class (es): 2 ]

Reference:

IWC-2500-1 Examination. Categories: C-C i Item Numbers: C3.20

== Description:== Alternative rules for the Inservice Inspection of Inaccessible Welds on Welded Attachments Component Number (s): Unit 1 Welds: l

1FW-06-33, *lFW-10-29, *1FW-11-29, *1FW-12-29 i
1RH-03-37B,
1 RH- 0 4 - 73A,
1 RH- 0 5-21B, 1RH-07-25A 1
1RH-08-02A, 1RH-09-45A, *1SI-04-02A, *1SI-12-23A 1SI-26-09A Unit 2 Welds:
2FW-06-01,
2 FW 2 4, *2FW-11-25,
2 FW-12-2 5
2RH-04-03, *2RH-05-35, *2RH-08-03, *2RH-09-28
2SI-04-03, *2SI-12-19A, *2SI-26-03

(*) denotes welds selected for inspection during the interval. CODE REQUIREMENT Subsection IWC, Table IWC-2500-1, Examination Categ ry C-C, Item C3.20 requires! surface examination of the Integrally Welded Attachments to Piping (Figure IWC-2500-5). BASIS FOR RELIEF Pursuant to 10 CFR 50.55a (a) (ii), relief is requested on the basis that compliance with the Code requirement would result in hardship or unusual difficulty without compensating increase in the level of quality and safety. Some penetrations at Braidwood were originally designed where one of the integral attachment welds is inside the flued head penetration assembly, thus making the welds inaccessible for inservice inspection. Access from outside of the closed end of the flued head penetration assembly for examiners is prohibited by the integral attachment. Access from the open end of the penetration is severely restrained due to geometry and clearance. See Attachments 1, 2, 3, 4 and 5 for penetration ' details. The integral attachment weld is set back some distance inside the flued head penetration assembly and the clearance between the pipe and penetration sleeve is small. See Table 1 on Attachment 6. To' satisfy the Code requirement to perform a surface examination of this weld, modification to the flued head penetration assembly and/or piping to allow access would be required. Braidwood would incur significant engineering and installation costs to perform such a modification without a compensating increase in tne level of quality and safety te justify such modifications. PROPOSED ALTERNATE PROVISIONS When a weld is. scheduled for inspection, a surface examination of the accessible weld on the exposed outside surface of the penetration will be performed. In conjunction with the above proposed-alternative technique, the periodic VT-2 I 3-64

-., -- -.... -.... _ _. - - - - - - - - - -... - - - ~ _... -... -. _ . _ - - ~.... - _........ - -. Revision 0 RELIEF REQUEST NR-26 (cont.) examinations in accordance with the requirements of ASME Section XI, Table IWC-2500-1, Examination Category C-H will provide reasonable assurance of continued structural integrity of the piping systems. APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval. APPROVAL STATUS Pending NRC review. F i 1 l l i I I l l l 1 4 3-65 i I.

1 i ) Revision 0 NR-26 ATTACHMENT 1 (Drawing not to scale) l CONTAINMENT WALL 13" r Tcw v See Detail A [ PRESSURE RETAINING PIPE v,r ts INACCESSIBLE WELD pg M: ACCESSIBLE WELD ANCHOR l PLATE.- 1/2" TYPICAL TYPE 2 PENETRATION CONTAINMENT WALL 13" [ I Tcw v' See Detail A PRESSURE RETAINING PIPE v,r ts INACCESSIBLE WELD pg ACCESSIBLE WELD ' ANCHOR. I PLATE 1/2" TYPICAL TYPE 3 PENETRATION 3-66

7.. ~ _..... -. - ..~.- -...- -..- -.... - - -.... i i l Revision 0 l NR-26 I ATTACHMENT 2 (Drawing not to scale) l i l PENETRATION ATTACHMENT I l i l T, I I i 1 45 1 1 An-4 i PROCESS PIPE WALL T. 1 i l DETAIL A I t 1 i i. J l i i 3-67 .v--

d i Revision 0 NR-26 ATTACHMENT 3 (Drawing not to scale) 4 i, t

CONCRETE WALL:

]' 6" I L " pC 3' See Detail B [ PRESSURE RETAINING PIPE ,\\ v, INACCESSIBLE WELD mn FHbbbUrk HblAINING FIFb 1rir ACCESSIBLE WELD '; ANCHOR. PLATE-TYPICAL PENETRATION (DETAIL 27) 3-68

4 4.LP_o 2 .-n.a. - - ._J a l Revision 0 NR-26 ATTACHMENT 4 i (Drawing not to scale) l i i - CONCRETE: WALL-4" or 6" F .l 3' \\ I See Detail B PRESSURE RETAINING PIPE vv \\ t INACCESSIBLE WELD Dg ACCESSIBLE WELD 7 TYPICAL PENETRATION (DETAIL 4) 3-69

- ~. _. -. -. i l l Revision 0 l NR-26 ) l ATTACHMENT 5 l (Drawing not to scale) t i l i I i I l l PENETRATION ATTACHMENT T, i j 45 I t i 1 I i t i PROCESS PIPE WALL T. 1 l DETAIL B 3-70

Revision 0 NR-26 ATTACHMENT 6 r TABLE 1 Weld ID Pipo Pipe Attachment Thickness Penet. Penet. Sizt Thickness Thickness Concrete size Clearance Te Ta Ten Pc TYPE 2 PENETRATIOAS (ATTACHMENT 1) 1RH-04-73A 12" 0.375" 2.0" 3' 6" 24" 4.9" 1RH-08-02A 12" 0.375" 2.0" 3' 6" 24" 4.9" 1SI-04-02A 8" 0.906" 2.0" 3' 6" 24" 7.0" 1SI-12-09A 12" 1.125" 2.0" 3' 6" 24" 4.9" 2RH-04-03 12" 0.375" 2.0" 3' 6" 24" 4.9" i 2RH-08-03 12" 0.375" 2.0" 3' 6" 24" 4.9" 2SI-04-03 8" 0.906" 2.0" 3' 6" 24" 7.0" 2SI-12-19A 12" 1.125" 2.0" 3' 6" 24" 4. 2, " TYPE 3 PENETRATIONS (ATTACHMENT 1) 1 1FW-06-33 6" 0.432" 2" 4 ' 6" 16" 3.8" 1FW-10-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 1FW-11-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 1FW-12-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 1SI-26-09A 8" 0.906" 2" 3' 6" 24" 7.0" 2FW-06-01 6" 0.432" 2" 4 ' 6" 16" 3.8" 2FW-10-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 2FW-11-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 2FW-12-29 6" 0.432" 2" 4 ' 6" 16" 3 8" 2SI-26-03 8" 0.906" 2" 3' 6" 24" 7.0" DETAIL 4 PENETRATIONS (ATTACHMENT 4) 1RH-03-37B 8" 0.375" 1* 3' 0" 18" 4.3" 1RH-07-25A 8" 0.375" 1" 3' 0" 18" 4.3" DETAIL 27 PENETRATIONS (ATTACHMENT 3) 1RH-05-21B 8" 0.375" 1" 3' 0" 20" 5.2" IRH-09-45A 8" 0.375" 1" 3' 0" 20" 5.2" 2RH-05-35 8" 0.375" 1" 3' 0" 20" 5.2" 2RH-09-28 8" 0.375" 1" 3' 0" 20" 5.2" 3-71

..... _.~.- - - - -. ~ - ~ _. _ _.- - -.- ~. ~. =. -. - ~ ~. -...- f A j Revision 0 RELIEF REQUEST NR-27 1 i 3 COFPONENT IDENTIFICATION 4 i I code Class (es): 1 1

Reference:

IWB-2500-1 1 j Examination Categories: B-D Item Numbers: B3.90

== Description:== Alternative rules for the Inservice Inspection of Reactor Vessel Nozzle to Vessel Welds i Component Number (s): Unit 1 Welds: 4 1RV-01-006, 1RV-01-009, 1RV-01-010, 1RV-01-013 1 E Unit 2 Welds i 2RV-01-006, 2RV-01-009, 2RV-01-010, 2RV-01-013 1 5 1-CODE REQUIREMENT j Subsection IWB, Table IWB-2500-1, Examination Category B-D, Item B3.90 requires essentially 100% volumetric examination of the region described in Figure IWB-2500-7 for Reactor Vessel nozzle-to-vessel welds. 4 BASIS EUR RELIEF Relief is requested on the basis that compliance with the Code requirement would result in hardship or unusual difficulty without compensating increase in the level i of quality and: safety. I i All of the welds contained in the Reactor Vessel are examined using remotely i operated underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concerns due to the high radiation i levels in these areas. The nozzle-to-vessel welds on the outlet (Hot Leg) nozzles are constructed with an integral extension which partially obstructs the circumferential scan for reflectors transverse to the weld (Reference Attachment 1). i This interference limits the exam aggregate coverage obtained for the weld and j adjacent base metal to about 81% instead of the Code required essentially 100% examination coverage. Strict ASME Section III quality controls were used when designing, fabricating and i installing these welds. In addition, these welds were volumetrically examined during Preservice inspections (PSI) with no rejectable irregularities found. The probability of a flaw occurring only in one of the areas not being examined is extremely small. Most future indications of significant size will be found by the {. examination of the weld as it currently exists. 4 Comed has recently performed this inspection at its Byron Station. Braidwood is of the same design and will deploy the same underwater inspection equipment as that used by the Byron Station. The above aggregate coverage is what was seen during the Byron Station inspection and is what would be expected for the Braidwood Station. j PROPOSED ALTERNATE PROVISIONS 2 The Reactor Vessel outlet (Hot Leg) nozzle welds will be examined to the fullest i extent practical using the available underwater volumetric inspection techniques. The periodic VT-2 examinaticns in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system i 3-72 e i

Revision 0 RELIEF REQUEST NR-27 (cont.) 1 monitoring requirements specified in the Technical Specifications will provide i reasonable assurance of continued structural integrity of the Reactor Vessel. 1 APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval. APPROVAL STATUS Pending NRC review. ) l l l 3-73 l

_. _ ~ 4 Revision 0 NR-27 ATTACHMENT 1 (Drawing not to scale) i +- 11.12 5" -+ (- .) Represents Code Volume BMT/2 Q.......... n BMT/2 R=BMT/2 Ti a \\// o a l as i 1 1 HORIZONTAL SECTION VIEW 41 55" .............2.8.,.97.,.......................................................... 57.75" VERTICAL SECTION VIEW i r ir /\\ w7 pd d R=BMT/2 BMT/2 _....... {. + 11.12 5 " --+ BMT/2 i Reactor Vessel Outlet Nozzle Note: BMT is abbreviation for Base Metal Thickness 3-74

_m l Revision 0 RELIEF REQUEST NR-28 i COMPONENT IDENTIFICATION i Code Class (es):' 1 )

Reference:

IWB-2500-1 j Examination Categories: B-D j Item Numbers: B3.100 1

== Description:== Alternative rules for the Inservice Inspection of Reactor Vessel Nozzle Inner Radius Sections (IRS) Component Number (s): Unit 1 Welds: 1RV-01-015, 1RV-01-016, 1RV-01-019, 1RV-01-020 Unit 2 Welds: 2RV-01-015, 2RV-01-016, 2RV-01-019, 2RV-01-020 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-D, Item B3.100 requires essentially 100% volumetric examination of the inner radius region described in ) Figure IWB-2500-7 for Reactor Vessel nozzle inner radius sections. BASIS FOR RELIEF Relief is requested on the basis that compliance with the Code requirement would result in hardship or unusual difficulty without compensating increase in the level of quality and safety. All of the inner radius sections of the vessel nozzles are exanined using remotely operated underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concerns due to the high radiation i levels in these areas. Examination of the inner radius sections of the inlet (Cold j Leg) nozzles is limited by the nozzle geometry (Reference Attachment 1 and 2). The design of the underwater volumetric inspection equipment is unable to cover the radius area where it transitions from the shell into the nozzle bore. This geometry ' limits the volumetric exam aggregate coverage obtained for the nozzle inner radius section to about 68% instead of the essentially 100% Code required exam volume. 1 The susceptibility of a PWR Reactor Vessel to flaw mechanisms within the inner ) radius regions is low. If a flaw were to develop in this region, the flaw would be j expected to initiate at or near the ID surface. A visual inspection of the ID 1 surface will detect early signs of such flaws before the flaws would propagate through-wall. Comed has recently performed this volumetric inspection at its Byron Station. .Braidwood is of the same design and will deploy the same underwater inspection equipment as that used by the Byron Station. The above aggregate coverage is wtat was seen during the Byron Station inspection and is what would be expected fe. che Braidwood Station. PROPOSED ALTERNATE PROVISIONS The Reactor Vessel inlet (Cold Leg) nozzle inner radius section will be examined to the fullest extent practical using the available underwater volumetric inspection technique. Also, a VT-1 of the nozzle inner radius area will be performed from the 3-75

Revision 0 interior of the Reactor Vessel using underwater camera equipment. In conjunction with the above proposed alternative technique, the periodic VT-2 examinations in RELIEF REQUEST NR-28 (cont.) accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structural integrity of the Reactor Vessel. APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval. APPROVAL STATUS Pending NRC review. l \\ 3-76 i

Revision 0 NR-28 i ATTACHMENT 1 a, (Drawing not to scale) 3 +- 11.12 5 " -+ o j BMT/2 1 Q ------.... T/2 (- - ) Represents Code Volume ~ \\// a a 6 2 See Detail A HORIZONTAL SECTION VIEW l ......... - -.. 2 7 4 7." .....--...........----------------------...--.....----....-----.00"------- 61 j VERTICAL SECTION VIEW ) I i r 1/2" /\\ f 6 R=BMT/2 BMT/2 ....---._ { +- 11.12 5 "

BMT/2 i

Reactor Vessel Inlet Nozzle Note: BMT is abbreviation for Base Metal Thickness 3-77

Revision 0 NR-28 ATTACHMENT 2 (Drawing not to scale) ) Approximate Cire. Scan Coverage 1/2" g-XcCC Detail A 3-78

___._..m. f Revision 0 RELIEF REQUEST NR-29 i COMPONENT IDENTIFICATION Code Class (es): 1 i i

Reference:

IWB-2500-1 Examination Categories: B-A j Item humbers: Bl.11, Bl.21, Bl.30 l I

== Description:== Alternative ruleb for the Jaservice Inspection of Reactor Vessel Circumferential Shell Welds, Lower Head circumferential Weld and Shell to Flange Weld. j k Component Number (s): Unit 1 Welds: 1RV-01-003, 1RV-01-004, 1RV-01-005, 1RV-02-001, I 1RV-02-002 Unit 2 Welds: 2RV-01-003, 2RV-01-004, 2RV-01-005, 2RV-02-001, l 2RV-02-002 [ CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-A, Items Bl.11, Bl.21 and Bl.30 requires volumetric examination be performed on the above Reactor Vessel Welds. These volumetric examinations are to be performed in accordance.with IWA-2232. IWA-2232 states that the inspections "shall be conducted in accordance with Article 4 of Section V", and amended by Section XI. Braidwood is currently committed to the Summer 1983 Addenda of Section V and XI. i BASIS FOR RELIEF j i Relief is requested on the basis that the proposed alternative to the described i requirements or portions thereof will provide an acceptable level of quality and safety. The Electric Utility industry has developed a program to qualify ultrasonic inspection techniques. This program, Performance Demonstration Initiative (PDI), is designed to meet the intent of Appendix VIII of the ASME Code, Section XI, 1992 Edition through 1993 Addenda. This program, PDI, used a variety of test blocks to evaluate transducer designs, scanning requirements and flaw sizing techniques. Braidwood has contracted with Framatome Technologies to use the URSULA manipulator to perform the 10 Year Ultrasonic (UT) Reactor Vessel inspections. Framatome Technologies has qualified techniques at the Performance Demonstration Initiative which meet the intent of' Appendix VIII of the ASME Code, Section XI, 1992 Edition through 1993 Addenda requirements. The techniques qualified by Framatome Technologies at PDI are summarized as follows. Flaw detection using the Appendix VIII PDI qualified technique would be accomplish by scanning in two scan directions, j one parallel and one perpend3cular to the weld. Flaw sizing using the Appendix VIII PDI qualified technique would be accomplished by scanning in four opposing scan { directions, two parallel and two perpendicular to the weld, when feasible, j PROPOSED ALTERNATE PROVISIONS 4 Braidwood proposes to use underwater volumetric inspection techniques which have been demonst;ated and qualified at the Performance Demonstration Initiative I Qualification to meet the intent of the tules of Appendix VIII of the ASME Code, 3-79

-.... - _ ~ ~.. - - -.. . - - - -.. -... -.. -... -.... ~... ~,. ... -. -.... - ~. - - - - I Revision 0 3 RELIEF REQUEST NR-29 (cont.) i f Section XI, 1952 Edition through 1993 Addenda in place of the currently required Section XI, Summer 1983 Addenda techniques. APPLICABLE TIME PERIOD ) This relief request will be required for the First Ten Year Inspection Interval. e i APPROVAL STATUS Pending NRC review. 'I t s I k I i 8 h I i l 'i I t 1 k I I i i 4 t 1 3-80 I ~ .. -.... ~..

Revision 0 i RELIEF REQUEST: NR-30 COMPONENT IDENTIFICATION Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-H Item Numbers: C7.30, C7.40, C7.70, and C7.80 l i

== Description:== Alternate Rules to Table IWC-2500-1, Category C-H: ~ Pressure Testing of Containment Penetration Piping with Attached Nonclassed Piping CODE REQUIREMENT ASME Section XI, Table IWC-2500-1, Examination Category C-H, ):equires the performance of a visual VT-2 examination during a system pressure test on Code Class 2 pressure retaining components. Note 7 of this table states, "The pressure i boundary includes only those portions of the system required to operate or support the safety system function _ up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required." ( BASIS FOR RELIEF t Pursuant to 10 CFR 50.55a (a) (3) (i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety. Also pursuant to 10 CFR 50.55a (a) (3) (ii), on the basis that compliance with the specified reqairements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. l Specifically, Braidwood Station requests relief to perform 10 CFR 50 Appendix J leakage testing in lieu of the pressure test required by ASME Section XI, Table IWC-2500-1, Examination Category C-H on the Code Class 2 Containment Penetration piping with attached nonclassed piping. I L The appliceble components are piping lines and valves which are portions of non- [ safety related systems that penetrate the primary reactor containment. At each l containment penetration, the process pipe is classified code Class 2 and provided with isolation valves that'are either locked shut during normal operation, capaole of automatic closure, or :apable of remote closure to support the containment safety f function. The balance of piping outside the isolation valves is non-code and therefore outside the scope of the ASME Boiler & Pressure Vessel Code, Section XI. These components perform no other safety function. The only reason that the penetration piping is classified as Class 2 is because of its function as part of the containment pressure boundary. The remaining portion of the system is non nuclear related and the integrity of the system in relation to its primary function is not within the scope of Section XI. Since containment integrity is the only safety.related function, it is logical to test the Class 2 penetration portion of the system to the Appendix J criteria. } The primary reactor containment integrity, including all containment penetrations, is periodically verified by performing leakage tests in accordance with 10 CFR 50, Appendix J. The Appendix J test frequency provides assurances that the containment i pressure boundary is being maintained at an acceptable level while monitoring for l, -deterioration of seals, valves and piping. If a pipe existed with a through wall flaw, the isolation valves located on both sides of the containment wall would prevent any release outside containment. Multiple through-wall flaws or leakage 3-81

~ l l i Revision 0 RELIEF REQUEST NR-30 (cont.) paths occurring simultaneously inside and outside of containment between the isolation valves in a pipe segment is unlikely. Each of the Code Class 2 lines and their associated isolation valves are tested during an Appendix J 1eakage test at a pressure not less than 44.4 psig (Peak calculated containment pressure). The l Appendix J leakage tests are performed at intervals in accordance with the requirements of the Braidwood Technical Specifications. Performance of these Appendix J leak tests will verify the integrity of the subject Code Class 2 lines and valves at the Containment penetrations. The performance of ASME Section XI, Examination Category C-H pressure tests on these same lines will provide little, if any, additional verification of primary reactor containment integrity and impose a burden of duplicate testing. Duplicate testing results in a significant increase in total amount of work force and radiological exposure without a compensating increase in the level of quality _. safety. Per the preceding information, Braidwood Station requests relief to use the Appendix l J test in lieu of the ASME Section XI requirements for pressure testing these Code l Class 2 containment penetration components on the basis that:

1) The current Code l

requirement results in a hardship or unusual difficulty without a compensating increase in the level of quality and safety and 2) The Proposed Alternate Provision. 1 provides an adequate level of quality and safety. The proposed alternative is consistent with the requirements of Code Ca se N-522. PROPOSED ALTERNATE PROVISIONS Braidwood Station will perform 10 CFR 50, Apnendix J leakage tests as an optional alternative to the Section XI required presst.re test on the subject primary reactor containment penetration piping and associatect valves, at intervals in accordance with Braidwood Station Technical Specificaticn 3/4.6. PERIOD FOR WHICH RELIEF IS REQUESTED Relief is requested for the first inspection interval. APPROVAL STATUS Pending NRC review. l l 3-82

~. ) I L ^ {. Revision 0 RELIEF REQUEST NR-31 1, j COMPONENT IDENTIFICATION Code Class (es): 1 d

Reference:

IWB-2500-1, Table IWB-2500-1 l l ' Examination Categories: B-A i J Item Number (s): Bl.ll

== Description:== Limited Volumetric Examination of Reactor Vessel Circumferential Shell Weld. Component Number (s): Unit 1 Weld: 1RV-02-002 Unit 2 Weld: 2 RV 002 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-A, Item Number B1.11 requires a '100% volumetric examination of the Reactor Vessel Circumferential Shell welds as detailed in Figure IWB-2500-1. BASIS FOR RELIEF j l Pursuant to 10 CFR 50.55a (a) (3) (ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Byron Station, daring Refuel Outage BIR07, conducted ultrasonic examinations of the Byron Unit 1 Reactor Vessel in April through June 1996. Framatome Technologies Inc. (FTI) was contracted to perform the examinations with their l state-of-the-art "URSULA" manipulator and their "ACCUSONEX" UT system. The Braidwood Reactor Vessel design is identical to the Byron Reactor Vessel design. Braidwood will deploy the same inspection equipment as that used at the Byron Station. 10CFR 50.55 a (g) (6) (ii) (A) revokes all relief requests with respect to volumetric examination coverages for welds specified in Item Bl.10. Portions of a previously granted relief request, NR-9, addressed limited exam coverage on the Braidwood Reactor Shell welds. During this exam at Byron, physical obstructions and geometry prevented UT coverage in excess of 90% of the required. volume. With having the same Reactor Vessel design as the Byron Station, the same coverage is expected for Braidwood Units 1 and 2. The examination of the Lower Shell Course-to-Dutchman weld, 1(2)RV-02-002, is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 4 inches above the weld (See Attachment 1). These lugs obstruct the automated UT inspection tool from examining the Code required volume of the weld below each lug (156' total for all 6 lugs). The FTI "URSULA" tool has a 6 degree movement arm and the physical size of the lugs and the " yaw" joint of the tool prevented scanning below the lugs back into the weld and surrounding base metal (See Attachment 2). All weld metal l can be examined between the lugs (204* total length between all 6 lugs). j Examinations for perpendicular and parallel reflectors covered areas i accounting for 57% of the weld metal and heat affected zone (HAZ). Sindla rly, 57% of the weld metal can be examined for transverse reflectors. 3-83 l F. iw w r-.w w - +

Revision 0 Compliance with the applicable Code requirements can only be accomplished by redesigning and refabricating the Reactor Vessel (s) and/or building a structure surrounding the vessel (s). Braidwood Station believes this course of action is a hardship without a compensating increase in the level of I quality and safety. The Code required volumetric examination will be completed to the maximum extent practical using ultrasonic examination techniques. The RPV examinations are conducted using an automated technique from the I.D. of the vessel. Access to allow inspection from the O.D. (shell side) of these welds 4 I is restricted due to the structural concrete surrounding the vessel. Reasonable assurance of the continued inservice structural integrity of the subject welds is achieved without performing a complete Code examination. The weld (s) have received visual examinations (VT-1 and VT-2) to visually verify s the integrity of the welds. Strict ASME Section III quality controls were used when designing, fabricating, and installing these welds. In addition, these welds were volumetrically examined during Preservice Inspections with no irregularities found. PROPOSED ALTERNATE PROVISIONS i The ultrasonic examination of the Braidwood Unit 1 and 2 reactor vessels will be performed to the maximum extent possible. No alternative volumetric examination is proposed to examine the areas not scanned due to obstructions or geometric constraints. VT-1 inspection will be conducted on ths weld (s) and HAZ(s) from the inside clad surface utilizing a submersible robot during the Refuel Outages. Additionally, a VT-2 examination during system pressure testing per Category B-P is performed on the Reactor Vessel each refueling outage to verify leaktight integrity of these welds. APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval. APPROVAL STATUS Pending NRC review. 3-84

Revision 0 NR-31 (Drawing not to scale) BMT ( 1 ) = 0. 62 5" L BMT(1)/2 r {1.50" '.24" i u 4h BMT(2)/2 BMT = Base Metal Thickness BMT(2)=5.625" Core Support Lug Deta2}. l l s 3-85 l m.

r Revision 0 i NR-31 (Drawing not to scale) t I l 1 l Inspectable Area Transducer Package p l i /N i N i ,N i i - l.. 'N(m' "N \\ 1: l // / \\ / w'N s 'g i 34' l ,/ g 13* i l 13* j \\ / i l ,/ i i i \\ l / Reactor Vessel Shell i 's l / i a i o Core Support Lug '( l j Obstructed Scan Area i Partial Plan View of Core Support Lugs j i i 1 3-86

. =... -. . ~ . ~. -. =. 4 Revision 4 Table 4.0-1 Summary of Technical Approaches and Positions (sheet 1 of 1) Note 1 The Main Steam Nozzle does not have a nozzle inner radius as described in Figure IWC-2500-4 (a) and (b) so no ultrasonic examination is required. Note 2 Code Case N-419 limiting exams (Code Category B-G-1) to those components f selected for Categories B-B, B-J, B-L-2, and B-M-2. j l Note 3 Code Case N-426 limiting exams (Code Category B-G-2) to those components selected for Categories B-B, B-J, B-L-2, and B-M-2. Note 4 Code Case N-401 which allows the use of digitally encoded eddy current data. Note 5 Augmented ultrasonic examinations will be performed on high energy pipe lines which have not been postulated for line breaks per Nuclear Regulatory Guice 1002 (Main Steam & Feedwater system). Note 6 Augmented Reactor Coolant Pump Flywheel exams will be performed per Nuclear Regulatory Guide 1.14 (ref. B/B UFSAR A1.14-3 and Technical Specification 3/4.4.10). Note 7 Ultrasonic re-examination will be performed on the Unit 1 Loop A upper i transition cone weld (lSG PC) per IWB-2420. Note 8 Augmented inspections of turbine rotors and discs will be performed per Westinghouse recommendations. Note 9 Ultrasonic examinations will be performed on accessible portions of the Reactor Vessel head welds (upper & lower) Note 10 Augmented ultrasonic examinations on 7 1/2s of the circumferential welds within the ECCS systems (RH, CV, CS, & SI) for the detection of IGSCC. Note 11 Ultrasonic re-examination will be performed on a Unit 2, Reactor Coolant I main loop weld (2RC-01-04) per IWB-2420. Note 12 Deleted (refer to Note 13) Note 13 Code Case N-408-2 exempts vessels with cumulative inlet and cumulative outlet pipe cross-sectional diameter which do not exceed 4" NPS. Note 14 Ancmented ultrasonic examinations will be performed on areas which may he subject to thermal stratification per NRC Bulletin 88-08. Note 15 Ultrasonic and surface examinations will be performed on four (4) Residual Heat exchanger nozzle to vessel welds in which indications were detected during A2R02 and AlR03 per IWC-2420. Note 16 Code Case N-416 defers system hydrostatic test required by IWA-4400 for repair or replacement of Class 2 piping. Note 17 Code Case N-498 allows a scheduled 10-year leakage test in lieu of a 10-year hydrostatic test. Note 18 Code Case N-356 allows certification period for Level III NDE Personnel to be extended to 5 years. Note 19 Code.. e N-460 provides alternate examination coverage for Class 1 and Class 2 welds. 4-2

_ = _.. Revision 0 i l NOTE 18 Braidwood will incorporate Code Case N-356 which extends the certification period for Level III NDE personnel to five (5) years. 4 1 I 1 i i e 4-27 a

- -... - = - -.... 1 i Revision 0 NOTE 19 j Braidwood will incorperate Code Case N-460 which provides Alternate Examination Coverage for Class 1 and 2 Welds. e f I J 8 I l ) f 4 4 } l i i \\ w J e 4-28

Revision 3 TABLE 5. 0-1 Comed Braidwood Nuclear Station Units 1&2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMAPES B-A Pressure retaining welds in reactor vessel i

Bl.10 Shell welds Bl.11 Circumferential 3 VOL NR-9,29,31 All welds Bl.12 Longitudinal 0 Bl.20 Head welds Bl.21 Circumferential 2 VOL NR-29 Note 9 Accessible Bl.22 Meridional 0 length of all welds Bl.30 Shell-to-Flange weld 1 VOL NR-29 1-weld Bl.40 Head-to-Flange weld 1 VOL and SUR NR-9 1-weld Bl.50 Repair welds Bl.51 Beltline Region 0 5-2

~._.- _._. __m. ....m Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1& 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS o

B-D Full penetration welds of nozzles in vessels s (Inspection Program B) Reactor Vessel B3.90 Nozzle-Vessel Welds 8 VOL NR-27 All Welds i i B3.100 Nozzle Inner Radius 8 VOL NR-28 all Welds Pressurizer B3.110 Nozzle-Vessel Welds 6 VOL NR-24 All Welds B3.120 Nozzle Inner Radius 6 VOL NR-24 All Welds Steam Generators B3.130 Nozzle-to-Vessel Welds O B3.140 Nozzle Inner Radius 8 VOL NR-21 All Welds (Unit 1 only) Heat Exchangers B3.150 Nozzle-to-Vessel Welds 0 B3.160 Nozzle Inner Radius 0 5-5

~ Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS B-E Pressure retaining partial penetration welds in vessels.

Reactor Vessel B4.10 Partial Penetration Welds B4.11 Vessel Nozzles 1 VT-2 Vent Pipe B4.12 Control Rod Drive 78 VT-2 Nozzles 25% of Nozzles B4.13 Instrumentation Nozzles 58 VT-2 25% of Nozzles Pressurizer B4.20 Heater Penetration 78 VT-2 Welds 5-6

Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL EXAMINATION ITEM APPROACH NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION RE:iARKS B-G-2 Pressure Retaining Bolting 5 2 in. dia.

Reactor Vessel B7.10 Bolts, Studs and Nuts 7 assemblies VT-1 2 RVLIS & 5 incore thermal-Pressurizer couples B7.20 Bolts, Studs and Nuts 1 Manway VT-1 Primary Steam Generator Manway B7.30 Bolts, Studs, and Nuts 8 Manways VT-1 Primary Heat Exchangers Manway B7.40 Bolts, Studs and Nuts O Piping B7.50 Bolts, Studs and Nuts 20(U2-21) VT-1 U1-11(U2-15) 5-13 t

Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL EXAMINATION ITEM APPROACH NUMBER OF EXAM RELIEF AND CRTEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS B-G-2 Pumps (Cont.)

B7.60 Bolts, Studs and Nuts 4 pumps VT-1 Note 3 1 Pump 6 B7.70 Valves 64 VT-1 Note 3 20 valves CRD Housings B7.80 Bolts, Studs and Nuts O l 5-14

Revision 3 TABLE 5.0-1 ~ Comed Braidwood Nuclear Station Units 1& 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS B-H Integral attachments for vessels Reactor Vessels B8.10 Integrally welded 0

attachments Pressurizer B8.20 Integrally welded 5 SUR NR-25 attachments Support Skirt Steam Generator B8.30 Integrally welded 0 attachments Heat Exchanger B8.40 Integrally welded 0 attachments 5-15 ~

Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS B-J Pressure Retaining Welds in Piping B9.10 Nominal Pipe Size > 4" Dia.

B9.11 Circumferential Welds 604(U2-554) VOL and SUR NR-2,6,7,18 25% of the B-B9.12 Longitudinal Welds O VOL and SUR B9.20 Nominal Pipe Size <4 in dia. B9.21 Circumferential Welds 139(U2-132) SUR 25% of the B-B9.22 Longitudinal O B9.30 Branch Pipe Connections 89.31 Nominal Pipe Size > 41n 11(U2-13) VOL and SUR 25% of the B-J Population B9.32 Nominal Pipe Size < 41n 54(U2-73) SUR 25% of the B-J Population B9.40 Socket Welds 840(U2-855) SUR 25% of the B-J Population l 5-16 [

" ^ - ~ ' ' " ~ ~~ Revision 3 TABLE 5.C-1 i Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 1 TECHNICAL EXAMINATION ITEM APPROACH NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS B-K-1 Integral attachments for Piping, Pumps, and valves Piping B10.10 Integrally Welded 7(U2-20)

SUR ' Attachments 3-U1 18-U2' Pumps B10.20 Integrally Welded 0 i Attachments Valves B10.30 Integrally Welded 0 Attachments

Revision 3 TABLE 5.O-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 2 TECHNICAL APPROACH EXAMZNATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER

_ DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS C-A Pressure Retaining Welds in Pressure Vessel C1.10 Shell Circ Welds 14 VOL Note 7, 13 4 Welds C1.20 Head Cire. Welds 6 VOL Note 13 2 Welds C1.30 Tubesheet-to-Shell 4 VOL Welds 1 Weld C-B Pressure Retaining Nozzle Welds in Vessels C2.10 Nozzles in Vessels 5 O Nominal Thickness C2.11 Nozzle-to-Shell Welds O C2.20 Nozzle Without Reinforcing Plates in vessel > in. Thickness C2.21 Nozzle-to-Shell Weld 16 SUR and VOL NR-23 Note 15 5 Welds C2.22 Nozzle Inside Radius 12 VOL NR-12 Note 1 1 Weld 5-22

.m., Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1& 2 ISI PLAN

SUMMARY

- CLASS 2 TECHNICAL EXAMINATION ITEM APPROACH NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS C-C Integral Attachments for Piping, Pumps and Valves Pres:mre Vessels C3.10 Integrally Welded 2

SUR Attachments Note 13 1 Welds Piping C3.20 Integrally Welded 43(U2-37) SUR NR-26 Attachments 37 Unit 1 Attachments 37 Unit 2 Pumps Attachments C3.30 Integrally Welded 14 SUR NR-15 Attachments 7 Attachments Valves C3.40 Integrally Welded 0 Attachments C-D Pressure Retaining Bolting > 2in. Diameter Pressure Vessels C4.10 Bolts and Studs 0 5-24

4 Revision 3 TALLE 5.0-1 Comed Braidwood Nuclear Station Units 1& 2 ISI PLAN

SUMMARY

- CLASS 2 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS C-F C5.30 Pipe Branch Connections (Cont.)

> 41n. Branch Pipe Size C5.31 Circumferential Welds 34(34-U2) SUR NR-1 U1=ll welds U2=10 welds C5.32 Longitudinal Welds 3-U1 SUR None C-G Pressure Retaining Welds in Pumps and Vlvs Pumps C6.10 Pump Casing Welds O valves C6.20 valves Body Welds O C-H All Pressure Retaining Components Pressure vessels C7.10 Pressure Retaining ALL VT-2 Components System Pressure Test 5-26 L ~ ~ ~

i Revision 3 .i TABLE 5.0-1 Comed Braidwood Muclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 2 TECHNICAL EXAMINATION ITEM APPROACH NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS l

C-H C7.20 Pressure Retaining ALL VT-2 Note 17 System Hydro-(Cont.) Components static Test Piping I C7.30 Pressure Retaining ALL VT-2 NR-30 Components System Pressure Test C7.40 Pressure Retaining ALL VT-2 NR-30 Note 17 System Hydro-l Components static Test i Pumps C7.50 Pressure Retaining ALL VT-2 Components System Pressure Test C7.60 Pressure Retaining ALL VT-2 Note 17 System Hydro-Components static Test Valves C7.70 Pressure Retaining ALL VT-2 NR-30 Components System Pressure Test C7.80 Pressure Retaining ALL VT-2 NR-30 Note 17 System Hydro-Components static Test i 5-27

Revision 3 TABLE 5.0-1 Comed Braiuwood Nuclear Station Units 1&2 ISI PLAN

SUMMARY

- CLASS 3 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBE7 DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS D-A Systems in Support of Reactor Shutdown Function D1.10 Pressure Retaining ALL VT-2 NR-23 Components Leakage Each Period Hydro-static Each Interval Dl.20 Integral Attachments 44(U2-35)

VT-3 U1=33 U2=27 Component Supports D1.30 Integral Attachments O Snubbers D1.40 Integral Attachments O Spring Type Supports D1.50 Integral Attachments O Constant Load Supports D1.60 Integral Attachments O Shock Absorbers D-B Systems in Support of ECCS Leakage Each D2.10 Pressure Retaining ALL VT-2 Components

Period, Hydro-static Each Interval 5-28 L

t Revision 3 TABLE 5.0-1 I Comed r Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 3 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS D-B D2.20 Integral Attachments 67(31-U2)

VT-3 (Cont.) U1=44, U2=15 Component Supports D2.30 Integral Attachments O Snubbers D2.40 Integral Attachments O Spring Type Supports D2.50 Integral Attachments O Constant Load Supports Shock Absorbers D2.60 Integral Attachments O D-C System in Support of RHR from the Spent Fuel Pool t D3.10 Pressure Retaining ALL VT-2 Components D3.20 Integral Attachments O Component f Supports D3.30 Integral Attachments 0 Snubbers i D3.40 Integral Attachments O Spring Type Supports D3.50 Integral Attachments O Constant Load i I Supports i '-l I

e

h Revision 3 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN

SUMMARY

- CLASS 3 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS D-C D3.60 Integral Attachments O

Shock Absorbers AUG Augmented High Energy Ul=314 VOL Note 5 All Welds Main Steam & Feedwater U2=330 AUG Reactor Coolant Pump 4 VOL and SUR Note 6 Each Period Flywheel Examination AUG Turbine Rotors Note 8 AUG Locations Subject to 7 VOL Note 14 Each Period Thermal Stratification AUG ECCS IGSCC U1-789 VOL Note 10 7 1/2% of the Examination U2-783 Welds Tables for subsection IWE have not been approved for use. Components identified in subsection IWF selected for examination shall be the supports of those components that are required to be examined under IWB, IWC, and IWD during the first inspection interval. 1 S-30 -}}

ML20129C401

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