ML20211G526
| ML20211G526 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 08/24/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211G523 | List: |
| References | |
| NUDOCS 9908310276 | |
| Download: ML20211G526 (19) | |
Text
~
)
7 h
NJ
[Q,
,k 1' NITE: ST?" 3 j
g ), g,j NUGLEAR REGULn 'OR'r COMMISSION L
g% s WA5mGTOR D C. MSN j
- ...+
SAFETY EVALUATION BY TH: OFFICE OF NUC". EAR REACTOR REGULATION 3
FOR RELIEF REQUEST 98-004 MCGUIRE NUCLEAR STATION. UNIT 1 DUKE ENERGY CORPORATION DOCKET NUMBER 50-369
1.0 INTRODUCTION
Inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code (ASME Code) Class 1,2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda, as required by Title 10 of the Code of Federal Reoulations (10 CFR) Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (0) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of qual;ty and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requiiaments, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of constraction of the components. The i
regulations require that inservice examination of components and system pressure tests conducted during the first ten year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to
]
the limitations and modifications listed therein. The Code of record for the McGuire Nuclear Station, Unit 1, second 10-year ISI intervalis the 1989 Edition of Section XI of the ASME Code.
2.0 EVALUATION By letter dated November 24,1998, Duke Energy Corporation (licensee) submitted its Second 10-Year IntervalISI Program Plan Request for Relief 98-004 for McGuire Nuclear Station, Unit 1.
9908310276 990824 PDR ADOCK 05000369 p
2-The Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the iaformation provided by the licensee in support of its S econd 10-Year Interval ISI Program Plan Request for Relief 98-004 for McGuire Nuclear Station, Uait 1. Based on the results of the review, the staff adopts the cor. tractor's conclusions and recommendations presented in the
- Technical Evaluation Report (TER) enclosed.
The information provided by the licensee in support of its alternative to the ASME Code requirements has been evaluated and the basis for disposition is documented below.
Reauest for Relief 98-004. Part A:
ASME Code,Section XI, Examination Category B-A, item B1.21, requires 100 percent volumetric examination, as defined by Figure IWB-2500-3, for Reactor Pressure Vessel (RPV) circumferential head welds.
Pursuant to 10 CFR 50.55a(g)(5)(iii), the Itcensee requested relief from the ASME Code coverage requirements for circumferential head Weld 1 RPV6-4468.
The staff has determined that complete examination of this weld is limited by component configuration (i.e., taper of the head) and adjacent physical obstructions (i.e., lifting lugs).
These restrictions limit access and make the ASME Code coverage requirement impractical for the subject head weld. To meet the ASME Code coverage requirement, design modifications would be necessary to provide access for examination. Imposition of the ASME Code requirements would be a significant burden on the licensee.-
The licensee has examined a significant portion (-75 percent) of RPV circumferential head Weld 1 RPV6-446B. In addition, other RPV welds are receiving volumetric examinations to the extent practical. These examinations provide reasonable assurance of the structuralintegrity of the subject components. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
Reauest for Relief No.98-004. Part B:
ASME Code,Section XI, Examination Category B-D, item B3.110 requires 100 percent volumetric examination, as defined by Figure IWB-2500-7, for pressurizer nozzle-to-vessel welds. Examination Category B-D, item B3.140, requires 100 percent volumetric examination, as defined by Figure IWB-2500-7, for steam generator nozzle inner radius (IR) sections.
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the ASME Code coverage requirements for the welds listed below.
. Component ID #
Item Number Coverage Pressurizer Surge Nozzle 1 PZR-10 B03.110.001 74.78 %
to lower Head Steam Generator Nozzle 1SGA-Inlet B03.140.001 75.93 %
Inside Radius Sections 1SGA Outlet 803.140.002 75.93 %
1SGD-Inlet B03.140.007 75.93%
1SGD-Outlet B03.140.008 75.93 %
1SGC-Inlet 803.140.005 82.28 % -
1SGC-Outlet 803.140.006 82.28 %
The staff has determined that complete examination of these welds is limited by nozzle configuration (i.e. nozzle bore size and vessel wall thickness). These restrictions limit access and make the ASME Code coverage requirements impractical for the subject head weld. To meet the ASME Code coverage requirements, design modifications would be necessary to provide access for examination. Imposition of the ASME Code requirements would be a significant burden on the licensee.
The licensee has examined a significant portion (275 percent) of each of the subject nozzle examination areas. The examinetions performed provide reasonable assurance of the structural integrity of the subject component. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
Reauest for Relief No.98-004. Part C:
ASME Code,Section XI, Examination Category B-F, item B5.10, B5.70 and B5.130 requires 100 percent volumetric and surface examination, as defined by Figure IWB-2500-8, for reactor vessel, steam generator, and piping dissimilar metal welds 4-inch nominal pipe size (NPS) or larger.
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the ASME Code coverage requirements for the welds listed in the table below.
Vessel item /ID #
Coveraae RPV B05.010.009 89.13 %
B05.010.010 89.13 %
B05.010.011 89.13 %
B05.010.012 89.13 %
Steam Generator B05.070.001' 48.61 %
B05.070.002*
48.61 %
Piping 805.130.002*
48.61 %
B05.130.003*
48.61 %
B05.130.017 82.76 %
B05.130.018 82.76 %
B05.130.019 82.76 %
B05.130.020 82.76 %
l l
- l Vessel l Item /ID #
l Coveraae l
- Item numbers B05.070.001, B05.070.002,805.130.002 and B05.130.003 were cut out and re-welded due to the Steam Generator beina replaced.
The staff has determined that complete examination of the subject welds was limited by component geometry (one-sided access) and material properties (attenuative austenitic grain structure). As supported by figures attached to the licensee's submittal, these restrictions limit access and make the ASME Code coverage requirements impractical for the subject dissimilar metal welds. To meet the ASME Code coverage requirements, design modifications would be necessary to provide access for examination. Imposition of the ASME Code requirements would be a significant burden on the licensee.
The licensee has volumetrically examined 48 percent to nearly 90 percent of each of the subject dissimilar metal welds, in addition to completing the surface examination. Furthermore, these welds are part of a larger population of dissimilar metal welds that are receiving complete examination. The examinations performed provide reasonable assurance of structuralintegrity of the subject welds. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
Reouest for Relief No.98-004. Part D:
ASME Code,Section XI, Examination Category B-J, item 89.31 requires 100 percent volumotric and surface examination, as defined by Figures IWB-2500-9, -10, and -11, for branch connection welds NPS 4 inches and larger.
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the ASME Code coverage requirements for the branch connection welds listed below.
ID Number item Number Coverage 1NC47-WN8 B09.031.004 47.4 %
1NC48 WN4A B09.031.004 77.9 %
1NC48-WN4B B09.031.004 77.9 %
1NCS2-WN6 B09.031.004 29.81 %
The staff has determined that complete examination of these welds was limited by component geometry, which prevents complete coverage from both sides of the weld. These restrictions limit access and make the ASME Code's examination requirements impractical for these welds.
To meet the ASME Code requirements, design modifications would be necessary to provide access for examination. Imposition of the ASME Code requirements would be a significant burden on the licensee.
The licensee has volumetrically examined a significant portion (approximately 78%) of two of the subject welds,47 percent of a third, and 30 percent of the fourth, in addition to completing the surface examination. Furthermore, these welds are part of a larger population of dissimilar
. metal welds that are receiving complete examination. The above examinations provide reasonable assurance of structuralintegrity of the subje-t welds. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.0 CONCLUSION
The staff concludes that certain inservice examinations are impractical and cannot be performed to the extent required by the ASME Code at McGuire Nuclear Station, Unit 1. The licensee's proposed alternatives provide reasonable assurance of structuralintegrity of the subject welds. Therefore, for all parts of Relief Request 98-004, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
Principal Contributor: T. McLellan Date: August 24, 1999 l
I
m TECHNICAL LETTER REPORT ON SECOND 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF 98-004 FOR DUKE ENERGY CORPORATION MCGUIRE NUCLEAR STATION. UNIT 1 DOCKET NUMBER: 50-369 1.
INTRODUCTION By letter dated November 24,1998, the licensee, Duke Energy Corporation, submitted Request for Relief 98-004, seeking relief from the requirements of the ASME Code,Section XI, for the McGuire Nuclear Station, Unit 1, second 10-year inservice inspection (ISI) interval. The Idaho National Engineering and Environmental Laboratory (INEEL) staff's evaluation of the subject request for relief is in the following section.
B.
EVALUATION The ir, formation provided by Duke Energy Corporation in support of Request for Relief No.98-004 from Code requirements has been evaluated and the basis for disposition is documented below. The Code of record for the McGuire Nuclear Station, Unit 1, second 10-year ISI interval, which is scheduled to end December 2002, is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.
2.1 Reauest for Relief No.98-004. Part A. Examination Cateoorv B-A. Item B1.21. Reactor Pressure Vessel Circumferential Head Weld 1 RPV6-446B Code Reauirement: Examination Category B-A, item 81.21, requires 100% volumetric examination, as defined by Figure IWB-2500-3, for RPV circumferential head welds.
Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code coverage requirements for circumferential head Weld 1 RPV6-4468.
Licensee's Basis for Proposed Alternative (as stated):
'During the ultrasonic examination of the Reactor Vessel Closure Head Weld 1RPV 6-4469 (Item Number B01.021.001) shown in Attachment 1, coverage of required examination volume could not be obtained. The examination coverage was limited to 74.87% Geometric limitations caused by the 18.5 degree taper and lifting lug interference resulted in limited coverage of the required volume. In order to achieve more coverage, the lif ting lugs would have to be relocated away from the weld and the taper would have to be redesigned to allow scanning completely over the weld.
"During the ultrasonic examination of the Reactor Vessel Closure Head Weld 1RPV 6-446B (item Number B01.021.001), coverage of required examination volume could not be obtained. The examination coverage was limited to 74.87%
Geometric limitations caused by the 18.5-degree taper and lifting lug interference
resulted in limited coverage of the required volume. In order to achieve more coverage the lifting lugs would have to be relocated away from the weld and the taper would have to be redesigned to allow scanning completely over the welo.
Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Cc poration. Reference Attachment 1 for scan coverage.
"Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWB-2500-3 could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity.
"The Reactor Vessel Closure Head Weld listed above is located on the McGuire Unit 1 Reactor Vessel. This weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e., embrittlement) associated with neutron bombardment. This weld was rigorously inspected by radiography and dye penetrant during construction and verified to be free from unacceptable fabrication defects, if a leak were to occur at the weld in question, the reactor coolant leakage calculation which is normally performed daily (and required by Technical Specifications to be performed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) would provide an early indication of leakage. The unidentified leakage specification is required by Technical Specifications to be performed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) would provide an early indication of leakage. The unidentified leakage specificatiosin Technical Specification 3.4.6.2 is 1 gpm. Several other indicatore such as containment radiation monitors EMF-38, -39, and -40, the containment floor and equipment sump levels, containment humidity instruments and the ventilation unit condensate drain tank level would provide early indication of weld leakage for prompt Operations and Engineering evaluation.
" Pursuant to 10 CFR 50.55a(g)(6)(i), granting this relief for the Reactor Vessel Flange to Upper Shell Weld (sic) will provide reasonable assurance of weld / component integrity, and is authorized by law. In addition, the requested relief will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility."
Licensee's Proposed Alternative (as stated):
"The use of radiography as an alternate volumetric examination for all the above listed components is not practical due to component thickness and geometric configurations. Other restrictions making radiography impractical are the physical barriers prohibiting access for placement of source, film, image quality indicator, 1
etc.
l "Since radiography is impractical, Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain rnaximum coverage to the extent practical of item Numbers referenced in Section i of this Request for Relief. No additional ultrasonic examinations are planned during the current interval for the welds referenced in Section 1 of this request.
2
"For the Class 1 components listed in Section 1 above, Duke Energy proposes to use the system leakage test to compliment the limited examination coverage. The Code requires (reference Table IWB 2500-1, item Number 815.) that a system leakage test be performed after each refueling outage. This test requires a VT s visual examination for evidence of leakage. This testing will provide adequate assurance of pressure boundary integrity.
Evaluation: The Code requires 100% volumetric examination of RPV circumferential head Weld 1 RPV6-446B. However, complete examination of this weld is limited by component configuration (i.e., taper of the head) and adjacent physical obstructions (i.e.,
lif ting lugs). As supported by figures attached to the licensee's submittal, these restrictions limit access and make the Code coverage requirement impractical for the subject head weld. To meet the Code coverage requirement, design modifications would be necessary to provide access for examination. Imposition of the Code requirement would result in an undue hardship on the licensee.
The licensee has examined a significant portion (~75%) of RPV circumferential head Weld 1 RPV6-4468. In addition, other RPV welds are receiving volumetric examinations to the extent practical. As a result, any existing patterns of degradation would have been detected and reasonable assurance of the continued structuralintegrity has been provided. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
2.2
_Reauest for Relief No.98-004. Part B. Examination Cateoorv B-D. Items B3.110 and B3.140. Pressurizer and Steam Generator Nozzle Welds Code Reauirement: Examination Category B-D, item B3.110 requires 100% volumetric examination, as defined by Figure IWB-2500-7, for pressurizer nozzle-to-vessel welds.
Examination Category B-D, Item B3.140, requires 100% volumetric examination, as defined by Figure IWB-2500-7, for steam generator nozzle inner radius (IR) sections.
Licensee's Code Relief Reouest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code coverago requirements for the welds listed below.
Component ID #
Item Number Coverage Pressurizer Surge Nozzle 1 PZR-10 B03.110.001 74.78 %
to lower Head Steam Generator Nozzle 1SGA-Inlet B03.140.001 75.93 %
Inside Radius Sections 1SGA-Outlet B03.140.002 75.93 %
1SGD-inlet B03.140.007 75.93 %
1SGD-Outlet 803.140.008 75.93 %
1SGC-Inlet B03.140.005 82.28 %
1SGC-Outlet B03.140.006 82.28 %
3
Licensee's Basis for Proposed Alternative (as stated):
"During the ultrasonic exarnination of the Pressurizer Surge Nozzle to lower Head 1PZR-10 (item Number B03.110.001) shown in Attachment 2, coverage of required examination solume could not be obtaintf The examination coverage was limited to 74.78%. Single-sided access caused by the nozzle geometry resulted in limited coverage of the required volume. In order to achieve more coverage, the nozzle would have to be redesigned to allow access from both sides.
"During the ultrasonic examination of the Steam Generator Nozzles (Nozzle inside Radius Section):
1 SGA-inlet (B03.140.001) 1 SGA-Outlet (803.140.002) 1 SGD-Inlet (803.140.007) 1 SGD-Outlet (803.140.008) shown in Attachment 3, coverage of required examination volume could not be obtained. The examination coverage was Umited to 75.93%. Limitations are caused by the ratio of the nozzle O.D. to the vessel thickness. When the nozzle O.D. is large in relation to the vessel thickness, less coverage can be obtained when scanning from the vessel side. See Note 6.
"During the ultrasonic examination of the Steam Generator Nozzles (Nozzle inside Radius Section):
1 SGC-Inlet (803.140.005)
I SGC-Outlet (B03.140.006) shown in Attachment 3, coverage of required examination volume could not be obtained. When the nozzle O.D. is large in relation to the vessel thickness, less coverage can be obtained when scanning from the vessel side. See Note 6.
" Note 6 Examinations from the nozzle boss and OD blend radius using compound angles, determining which angles to use, metalpaths to calibrate for an area of coverage is not accurate with manual calculations. Duke Energy Corporation is investigating the use of computer modeling to solve the limitation problems.
"During the ultrasonic examination of the Pressurizer Surge Nozzle to Lower Head 1PZR-10 (item Number 803.110.001), coverage of the required examination volume could not be obtained. The examination coverage was limited due to single sided access caused by the nozzle geometry resulting in limited coverage of the i
required volume. In order to achieve more coverage the nozzle would have to be I
redesigned to allow access from both sides. Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would I
create a considerable burden on Duke Energy Corporation. Reference Attachment 2 for scan coverage.
" Steam Generator (Nozzle inside Radius Sections):
1SGA-Inlet (B03.140.001) 1 SGA-Outlet (803.140.002) 1SGC-Inlet (803.140.005) 4
+
1 SGC-Outlet (803.140.006) 1 SGD-Inlet (B03.140.007) 1 SGD-Outlet (803.140.008) turing the ultrasonic examination of the Leam Generator Nozzle inside Radius Sections listed above, coverage of required examination volume could not be obtained, limitations are caused by the ratio of the nozzle O. D. to the vessel thickness. When the nozzle O. D. is large in relation to the vessel thickness, less coverage can be obtained when scanning from the vessel side. Examinations from the nozzle boss and O. D. blend radius using compound angles, determining which angles to use, metal paths to calibrate for and area of coverage is not accurate with manual calculations. Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Corporation.
" Duke Energy Corporation is investigating the use of computer modeling to solve the limitation problems. Reference Attachrnent 3 for scan coverage.
"Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWB-2500-74 could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity.
"The seven welds listed above are located within the reactor coolant loop. These welds are not exposed to significant neutron fluence and are not prone to negative material property changes (i.e., embrittlement) associated with neutron bombardment. These welds were rigorously inspected by radiography and dye penetrant during construction and verified to be free from unacceptable fabrication defects. If a leak were to occur at any of the welds in question, the reactor coolant leakage calculation which is normally performed daily (and required by Technical Specifications to be performed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) would provide an early indication of leakage. The unidentified leakage specification in Technical Specification 3.4.6.2 is 1 gpm. Several other indicators such as containment radiation monitors EMF-38,
-39, and -40, the containment floor and equipment sump levels, containment humidity instruments and the entilation unit condensate drain tank level would provide early indication of y Md leakage for prompt Operations and Engineering evaluation.
- Note: On the four exams that took place prior to the Steam Generator Replacement (ISGA-Inlet, ISGA-Outlet, ISGD-Inlet,1SGD-Inlet) 75.93% coverage was obtained. On the two exams that took place after the Steam Generator Replacement (1SGC-Inlet, ISGC-Outlet) 63.28% coverage was obtained.
"Pursuar.t to 10 CFA 50.55a(gno)(i), granting this relief for the welds listed uriuer Examination Category B-D Welds will provide reasonable assurance of weld / component integrity, and is authorized by law. In addition, the requested relief will not endanger life or property or the common defense and security and is 5
otherwise in the public interest giving due consideration to the buroen upon the licensee that could result if the requirements were imposed on the facility."
Licensee's Proposed Alternative (as stated):
"The use of radiography as an alternate volumetric examination for all the above listed components is not practical due to component thickness and geometric configurations. Other restrictions making radiography impractical are the physical barriers prohibiting access for placement of source, film, image quality indicator, etc.
"Since radiography is impractical, Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practin! of item Numbers referenced in Section 1 of this Request for Relief. No additional ultrasonic examinations are planned during the current interval for the welds referenced in Section 1 of this request.
"For the Class 1 components listed in Section 1 above, Duke Energy proposes to use the system leakage test to compliment the limited examination coverage. The Code requires (reference Table IWB-2500-1, item Number B15.) that a system leakage test be performed after each refueling outage. This test requires a VT-2 visual examination for evidence of leakage. This testing will provide adequate assurance of pressure boundary integrity.
Evaluation: The Code requires 100% volumetric examination of the subject nozzle-to-vessel weld and nozzle IR sections. However, complete examination of these welds is limited by nozzle configuration (i.e. nozzle bore size and vessel wall thickness). As supported by figures attached to the licensee's submittal, these restrictions limit access and make the Code coverage requirements impractical for the subject head weld. To meet the Code coverage requirements, design modifications would be necessary to i
provide access for examination. Imposition of the Code requirements would result m an undue hardship on the licensee.
The licensee has examined a significant portion (275%) of each of the subject nozzle examination areas. As a result, any existing patterns of degradation would have been detected and reasonable assurance of the continued structuralintegrity has been provided. Therefore,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
2.3 Reauest for Relief No.98-004. Part C. Examination Cateaories B-F. Items B5.10. B5.70 and 65.130. Dissimilar Metal Welds Code Reauirement: Examination Category B F, item B5.10, B5.70 and B5.130 require 100% volumetric and surface examination, as defined by Figure IWB-2500-8, for reactor vessel, steam generator, and piping dissimilar metal welds 4-inch NPS or larger.
Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code coverage requirements for the welds listed in the 6
~
table below.
IVessel item /ID #
I Coverace l
RPV B05.010.009 89.13 %
B05.010.010 89.13 %
B05.010.011 89.13 %
~
B05.010.012 89.13 4 Steam Generator B05.070.001
- 48.61 %
B05.070.002*
48.61 %
Piping 805.130.002*
48.61 %
B05.130.003*
48.61 %
B05.130.017 82.76 %
B05.130.018 82.76 %
B05.130.019 82.76 %
B05.130.020 82.76 %
- Item numbers 805.070.001, B05.070.002, B05.130.002 and 805.130.003 were cut out and re-welded due to the Steam Generator being replaced.
Licensee's Basis for Proposed Alternative (as stated):
"During the ultrasonic examination of the Reactor Vessel Head to UH1 Tube Welds:
1RPV 1-462A-SE (B05.010.009) 1RPV 1-4628-SE (805.010.010) 1RPV 1-462C-SE (805.010.011) 1 RPV 1-462D-SE (805.010.012) shown in Attachments 4 and 5, coverage of required examination volume could not be obtained. The examination coverage was limited to 89.13%.
See Note 7.
"During the ultrasonic examination of the Steam Generator Nozzle-to-Safe end butt welds:
1 SGA-Inlet (B05.070.001) 1SGA-Outlet (805.070.002) shown in Attachment 6, coverage of required examination volume could not be obtained. The examination coverage was limited to 48.61%.
See Note 7.
"During the ultrasonic examination of the Piping Dissimilar Metal Butt Welds (Steam Generator Safe End to Pipe):
1 NC1 F-1-2 (B05.130.002) 1 NC1 F-i-3 (B05.130.003) shown in Attachment 6, coverage of required examination volume could not be obtained. The examination coverage was limited to 48.61%.
See Note 7.
7
,?
"During the ultrasonic examination of tne Piping Dissimilar Metal Butt Welds (UH1
- Tube to End Cap Welds):
1 N11 FW-38-3 (B05.130.017) 1 N11 FW-38-2 (805.130.018) 1 N11 FW-38-1 (B05.130.019) 1 NI1 FW-38-4 (805.130.020) shown in Attachments 4 and 5, coverage of required examination volume could not be obtained. The examination coverage was limited to 82.67%
See Note 7.
Note 7
Material characteristics and single sided access caused by the component geometryprevents two beam path direction coverage of the examination volume.
"The most effective ultrasonic technique for the examination of dissimilar metal welds uses refracted longitudinal waves. 'he longitudinal wave is preferred as the austenitic weld metal and buttering when present create highly attenuative barriers to shear wave ultrasound, Longitudinal waves provide superior penetration and improved signal to noise ratio over shear waves. However, the longitudinal wave is affected by mode conversion when it strikes the inside surface of the safe end or pipe at any angle other than a right angle to the surface.
"The calculations below show that a 45' refracted longitudinal wave striking the inside surface of a pipe will produce a 22.9* refracted shear wave in addition to the normally expected 45* reflected longitudinal wave.
Sin = (sin 45" x V ) + Ve 3
= (0.707 x 0.123) +0.223 Where sin is the shear wave angle VS is the shear wave velocity of the stainless steel safe end/ pipe materialin inches /psec.
VL is the longitudinal wave velocity of the stainless steel safe end/ pipe materialin inches /psec.
"As shown in the graph below', the mode conversion process creates two sound beams of differing intensities reflecting off the inside surface. At incident angles greater than 30* the shear wave will predominate. However, the shear wave is attenuated and scattered by the austenitic weld metal and the layer of buttering.
The examination sensitivity is degraded to such an extent that any examination using the second sound path leg is meaningless. Therefore, the two beam path direction coverage requirement is impractical. In order to obtain the required two-Not included in this report.
8
beam path direction coverage, welds would have to be re-designed to allow scanning from both sides.
Reactor Vessel Head to UH1 Tube Welds 1RPV 1-462A-SE (B05.010.009) 1 RPV 1-462B-SE (805.010.010) 1 RPV 1-462C-SE (B05.010.011) 1RPV 1-462D-SE (805.010.012)
Piping Dissimilar Metal Butt Welds (UH1 Tube to End Cap Welds):
1N11FW 38-3 (805.130.017) 1 N11 FW-38-2 (B05.130.018) 1N11FW-38-1 (805.130.019) 1 N11 FW-38-4 (805.130.020)
"These eight Dissimilar Metal Welds are limited due to material characteristics and single sided access caused by the component geometry prevents two-beam path direction coverage of the examination volume. In order to obtain the required two-beam path direction coverage, these welds would have to be redesigned to allow scanning from both sides. Replacement or redesign of these Class 1 welds is not a viable alternative and would create an undue burden on Duke Energy Corporation. During the examination of these welds, techniques were utilized to obtain the maximum possible covemge. Reference Attachments 4 and 5 for scan coverage.
"Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWB-2500-8 could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and j
integrity.
"The reactor coolant system welds listed above are located on the reactor vessel closure head. These welds are not exposed to significant neutron fluence and are not prone to negative material property changes (i.e., embrittlement) associated with neutron bombardment. These welds were rigorously inspected by radiography and dye penetrant during construction and verified to be free from unacceptable fabrication defects. If a leak were to occur at any of the welds in question, the reactor coolant leakage calculation which is normally performed daily (and required by Technical Specifications to be performed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) would provide an early indication of leakage. The unidentified leakage specification in Technical Specification 3.4.6.2 is 1 gpm. Several other indicators such as containment radiation monitors EMF-38, -39, and -40, the containment floor and equipment sump levels, containment humidity instruments and the ventilation unit condensate drain tank level wouid provide =>arly indication of weld leakage for prompt Operations and Engineering evaluation.
Steam Generator Nozzle-to-Safe End Butt Welds:
1 SGA-Inlet-SE (B05.070.001) 9 i
4 1 SGA-Outlet-SE (805.070.002)
Piping Dissimilar Metal Butt Welds (Steam Generator Safe End to Pipe):
1 NC1 F-1-2 (805.130.002) 1 NC1 F-1-3 (805.130.003)
"These four Dissimilar Metal Butt Welds are limited due to matedal characteristics and single-sided access caused by the component geometry prevents two-beam path direction coverage of the examination volume. In order to obtain the required two-beam path direction coverage, these four welds would have to be re-designed to allow scanning from both sides.
)
1 "The Steam Generator Nozzle to Safe End Butt Welds (Weld Numbers 1SGA-INLET-SE and 1SGA-OUTLET-SE) are located on the inlet and outlet of the steam generators for the reactor coolant piping. The Steam Generator Nozzle to i
Safe End Weld geometry prevented obtaining greater than 90% volumetric examination coverage. The imposition of this requirement would create a considerable burden on Duke Energy Corporation. During the examination of these welds, techniques were utilized to obtain the maxirnum possible coverage.
Reference Attachment 6 for scan coverage.
{
)
"Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWB-2500-8 could not be met, the amount of coverage obtained for these examinations provided an acceptable level of quality and integrity. Furthermore, these welds were cutout and re-welded during the steam generator replacement (1 EOC11 outage). These new welds were performed by FTl and received a cornplete radiographic examination, which were also performed by FTI, and verified to be free from unacceptable fabrication defects. There is no safety significance to the lack of weld examination coverage for the previous cycle.
" Pursuant to 10 CFR 50.55a(g)(6)(i), granting this relief for the welds listed under Examination Category B-F Welds will provide reasonable assurance of weld / component integrity, and is authorized by law. In addition, the requested relief will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility."
Licensee's Proposed Alternative (as stated):
"The use of radiography as an alternate volumetric examination for all the above listed components is not practical due to component thickness and geometric configurations. Other. restrictions making radiography impractical. are the physical barriers prohibiting access for placement of source, film, image quality indicator, etc.
"Since radiography is impractical, Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of item Numbers referenced in Section 1 of this Request for Relief. No additional ultrasonic examinations are planned during the current interval for the 10
welds referenced in Section 1 of this request.
"For the Class 1 components listed in Section 1 above, Duke Energy proposes to use the system leakage test to compliment the limited examination coverage. TM Code requires (reference Table IWB-2500-1, item Number B15.) that a system leakage test be performed after each refueling outage. This test requires a VT-2 visual examination for evidence of leakage. This testing will provide adequate assurance of pressure boundary integrity."
Evaluation: The Code requires 100% surface and volumetric examination for dissimilar metal safe end welds. However, complete examination of the subject welds was limited by component geometry (one-sided access) and material properties (attenuative austenitic grain structure). As supported by figures attached to the licensee's submittal, these restrictions limit access and make the Code coverage requirements impractical for the subjcct dissimilar metal welds. To meet the Code coverage requirements, cesign modifications would be necessary to provide access for examination. Imposition c! the Code requirements would result in an undue hardship on the licensee.
The licensee has volumetrically examined 48 to nearly 90% of each of the subject dissimilar metal welds, in addition to completing the surface examination. Furthermore, these welds are part of a larger population of dissimilar metal welds that are receiving complete examination. As a result, any existing patterns of degradation would have been detected and reasonable assurance of the continued structuralintegrity has been provided. Therefore,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
2.4 Reauest for Relief No.98-004. Part D. Examination Cateaorv B-J. Item B9.31. Class 1 Branch Connection Welds Code Reauirement: Examination Category B-J, item B9.31 requires 100% volumetric and surface examination, as defined by Figures IWB-2500-9, -10, and -11, for branch connection welds NPS 4-inch and larger.
Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code coverage requirements for the branch connection welds listed below.
!ID Number item Number Coverage 1NC47-WN8 B09.031.004 47.4 %
1NC48-WN4A B09.031.004 77.9 %
1NC48-WN4B B09.031.004 77.9 %
1NC52-WN6 B09.031.004 29.81 %
Licensee's Basis for Proposed Alternative (as stated):
"During the ultrasonic examination of Weld Number 1NC47-WN8 (809.031.004) shown in Attachment 7, coverage of required examination volume could not be obtained. The examination coverage was limited to 47.4%. See Note 8 11
3 "During the ultrasonic examination of Weld Number 1NC48-WN4A (B09.031.005) i shown in Attachment 8. coverage of required examination volume could not be obtained. The examination coverage was limited to 77.9% See Note 8
'During the ultrasonic examination of Weld Number 1NC48-WN4B (809.031.006) shown in Attachment 9, coverage of required examination volume could not be obtained. The examination coverage was limited to 77.9% See Note 8 "During the ultrasonic examination of Weld Number 1NC52-WN6 (809.031.007) shown in Attachment 10, coverage of required examination volume could not be obtained. The examination coverage was limited to 29.81% See Note 8
" Note 8 Single sided access caused by the branch connection geometry prevents scanning from both sides of the weld.
"In both cases cast stainless steel characteristics mandate the use of refracted longitudinal waves. This type of ultrasonic wave produces mode conversion at the pipe inside surface, thus preventing the use of sound path distances beyond the first " leg. Therefore, coverage of the required examination volume in two-beam path directions is not practical. In order to obtain the required two beam path direction coverage, the branch connections would have to be re-designed at allow scanning from both sides of the weld over the required examination volume."
"During the ultrasonic examination of Weld Number 1NC47-WN8 (809.031.004) shown in Attachment 7, coverage of required examination volume could not be obtained. The examination coverage was limited due to single-sided access caused by the branch connection geometry that prevents scannin'g from both sides of the weld.
"During the ultrasonic examination of Weld Number 1NC48-WN4A (B09.031.005) shown in Attachment 8, coverage of required examination volume could not be obtained. The examination coverage was I;mited due to single-sided access caused by the branch connection geometry that prevents scanning from both sides of the weld.
"During the ultrasonic examination of Weld Number 1NC48-WN4B (809.031.006) shown in Attachment 9, coverage of required examination volume could not be obtained. The examination coverage was limited due to single-sided access caused by the branch connection geometry that prevents scanning from both sides of the
- weld.
{
"During the ultrasonic examination of Weld Number 1NC52-WN6 (809.031.007) shown in Attachment 10, coverage of required examination volume could not be obtained. The examination coverage was limited due to single-sided access caused by the branch connection geometry that prevents scanning from both sides of the j
weld.
i 12
)
7 i
{
"In order to obtain the required coverage these welds would have to be redesigned.
The 100% volumetric examination is impractical due to nozzle or weld material geometry, or branch piping interferences. Replacement or redesign of this piping Class 1 piping is not a viable alternative and would create an undue burden on
{
Duke Energy Corporation. During the exar,anation of these welds, techniques were utilized to obtain the maximum possible coverage. Reference Attachments 7 thru 10 for scan coverage.
"Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figures IWB-2500-9 thru -11 could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity.
"The reactor coolant system piping branch connection nozzle welds listed above are located on the reactor coolant loop piping. These welds are not exposed to significant neutron fluence and are not prone to negative material property changes (i.e., embrittlement) associated with neutron bombardment. These welds were rigorously inspected by radiography and dye penetrant during construction and verified to be free from unacceptable fabrication defects. If a leak were to occur at any of the welds in question, the reactor coolant leakage calculation which is normally performed daily (and required by Technical Specifications to be performed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) would provide an early indication of leakage. The unidentified leakage specification in Technical Specification 3.4.6.2 is 1 gpm. Several other indicators such as containment radiation monitors EMF-38, -39, and -40, the containment floor and equipment sump levels, containment humidity instruments and the ventilation unit condensate drain tank level would provide early indication of weld leakage for prompt Operations and Engineering evaluation.
Licensee's Proposed Alternative (as stated):
"The use of radiography as an alternate volumetnc examination for all the above listed components is not practical due to component thickness and geometric configurations. Other restrictions making radiography impractical are the physical barriers prohibiting access for placement of source, film, image quality indicator, etc.
"Since radiography is impractical, Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of item Numbers referenced in Section 1 of this Request for Relief. No additional ultrasonic examinations are planned during the current interval for the welds referenced in Section 1 of this request.
"For the Class 1 components listed in Section 1 above, Duke Energy proposes to use the system leakage test to compliment the limited examination coverage. The Code repres (reference Table iWB-2500-1, item Number 815.) that a system leakage test be performed after each refueling outage. This test requires a VT-2 visual examination for evidence of leakage. This testing will provide adequate j
assurance of pressure boundary integrity."
13
" Pursuant to 10 CFR 50.55a(g)(6)(i), granting this relief for the welos listed under Examination Category B-J will provide reasonable assurance of weld / component integrity, and is authorized by law. In addition, the requested relief will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the f acility."
Evaluation: The Code requires 100% surface and volumetric examination of the subject branch connection welds. However, complete examination of these welds was limited by component geometry, which prevents complete coverage from both sides of the weld. As supported by figures attached to the licensee's submittal, these restrictions limit access and make the Code's examination requirements impractical for these welds. To meet the Code requirements, design modifications would be necessary to provide access for examination, imposition of the Code requirements would result in an undue hardship on the licensee.
The licensee has volumetrically examined a significant portion (approximately 78%) of two of the subject welds,47% of a third, and 30% of the fourth, in addition to completing the surface examination. Furthermore, these welds are part of a larger population of dissimilar metal welds that are receiving complete examination. As a result, any existing patterns of degradation would have been detected and reasonable assurance of the continued structuralintegrity has been provided. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
C.
CONCLUSION The INEEL staff evaluated the licensee's submittal and concluded that certain inservice examinations cannot be performed to the extent required by the Code at McGuire Nuclear Station, Unit 1. Therefore, for all parts of Request for Relief 98-004, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
14