U-602566, Forwards Requests for Relief from Requirements of Section XI of ASME Boiler & Pressure Vessel Code,1980 Edition Through Winter 1981 Addenda,W/Re Performance of non-destructive Exam of Welds

Forwards Requests for Relief from Requirements of Section XI of ASME Boiler & Pressure Vessel Code,1980 Edition Through Winter 1981 Addenda,W/Re Performance of non-destructive Exam of Welds

ML20117K210
Person / Time
Site:	Clinton
Issue date:	06/05/1996
From:	Lyon M ILLINOIS POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L30-96(06-05)LP, L30-96(6-5)LP, U-602566, NUDOCS 9606100150
Download: ML20117K210 (41)

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P.O. Box 678 Chnton, IL 61727 Tel 217 935-8881 ILLIN9IS P9WER u e 2566 L30-96(06 -05 )LP 8E.100c June 5, 1996 Docket No. 50-461 Document Control Desk 10CFR50.55a Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Submittal of Nine ASME Section XI Relief Requests Regarding Performance ofNon-Destructive Examinations for the First 120-Month Inservice Inspection Interval for Clinton Power Station

Dear Madam or Sir:

Attached for NRC review and approval are nine requests for relief from the requirements of Section XI of the American Society ofMechanical Engineers (ASME)

Boiler and Pressure Vessel Code,1980 Edition through Winter 1981 Addenda, with regard to the performance of non-destructive examination of welds and components for the first 120-month inservice inspection interval for Clinton Power Station (CPS). Details concerning the applicable ASME Section XI Code requirements, the affected components, and the basis and justification for the requested relief are provided in each of the attached relief requests (Attachments 1 through 9 of this letter). In accordance with NRC guidance, a relief request has been prepared for each ASME Section XI Examination Category. The format of each attached relief request is based on the format that has been used to date in the CPS Inservice Inspection Program manual. A brief description of each relief request (along with the corresponding attachment number to this letter) is provided below:

Relief Request No.

Attachment No.

Descriotion 4005 1

This is a request to allow a reduction in the required volumetric examination volume for reactor pressure vessel (RPV) head-to-flange weld CH-C-2 (Code Class 1, 100038 Examination Category B-A, Item Bl.40). The RPV head flange configuration / geometry limits the ability to fully scan this weld for ultrasonic examination.

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Page 2 Relief Request No.

Attachment No.

Descriotion 4006 2

This is a request to allow a reduction in the required volumetric examination volume of 29 RPV nozzle-to-shell welds (Code Class 1 Examination Category B-D, l

Item B3.90). The RPV nozzle configuration / geometry limits the ability to fully scan these welds for ultrasonic ~

examination.

4007 3

This is a request to allow Illinois Power (IP) to select a different weld in the feedwater system for surface and volumetric examination in lieu of weld 1-FW-2-5 which is required to be examined per Note 1(b) of Table IWB-2500-1 (Code Class 1, Examination Category B-J, Item B9.11). Examination of weld 1-FW-2-5 (a pipe-to-valve i

weld) is obstructed by a large hanger (guide support).

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4008 4

This is a request to allow a reduction in the required surface examination area for certain welds associated with integral attachments on piping in the Main Steam and Reactor Recirculation system (Code Class 1, Category -

B-K-1, Item B10.10), and to permit use of ASME Section XI Code Case N-509 which would allow a i

sample of attachment welds to be examined in lieu of examining all of the welds for shock suppressor lugs attached to the bowls of the "A" and "B" Reactor Recirculation pumps (Code Class 1, Category B-K-1, Item B10.20). Full examination of the piping integral attachment welds per the Code would require the undesirable and difficult removal / disassembly of piping supports.

4009 5

This is a request to require visual (VT-3) inspection of internal surfaces of the reactor recirculation pumps (Code Class 1, Examination Category B-L-2, Item B12.20) and certain valves (Code Class 1, Examination Category B-M-2, Item B12.50) only when opened for maintenance (versus being required to be done regardless of any maintenance performed). This will potentially prevent the unnecessary disaasembly of a large pump and up to 11 valves in the next refueling outage. This request is based on allowing IP to implement the applicable requirement of the 1989 edition of ASME Section XI (vice the edition to which IP is currently committed for the first ten-year interval).

U-602566 Page 3 Relief -

Request No.

Attachment No.

Descriotion 4010 6

This is a request to allow IP not to perform required volumetric examinations of certain RPV circumferential and vertical welds (Code Class 1, Examination Category B-A, Items Bl.11 and Bl.12). This request is based on allowing IP to adopt, as an alternative to the Code, the requirements of EPRI Topical Report TR-105697 (submitted to the NRC in September,1995). A reduction i

in the number of required weld examinations would yield considerable savings in resources, dollars, and exposure.

4011 7

This is a request to allow a reduction in the required examination volume of head-to-vessel weld HEA-1 on RHR Heat Exchanger A (Code Class 2, Examination Category C-A, Item C1.20). Permanently i

installed / welded lifting lugs limit the ability to fully scan this weld for ultrasonic examination.

4012 8

This is a request to allow a reduction in the required surface examination area for four integral attachment welds on the Reactor Core Isolation and Cooling pump casing (Code Class 2, Examinatwn Category C-C, Item C3.30). Since the associated attachments are utilized to mount the pump to its pedestal, lower portions of these attachment welds are not accessible for examination without removing the pump from its pedestal.

4013 9

This is a request to allow a reduction in the required surface examination area of weld RHR-A-2 on the casing of RHR pump "A" (Code Class 2, Examination Category C-G, Item C6.10). A permanently installed instrument line partially obstructs examination of this weld.

Inservice inspection of ASME Code Class 1,2, and 3 components is required to be performed in accordance with Section XI of the ASME Code pursuant to 10CFR50.55a(g), except where specific written relief has been granted by the Commission. Per 10CFR50.55a(g)(5), if a licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the NRC in support of that determination and a request made for relief from the ASME Code requirement. Further,10CFR50.55a(a)(3) states that alternatives to the requirements of 10CFR50.55a(g) may be used, when authorized by the NRC, if the proposed alternatives would provide an acceptable level of quality and

U-602566 Page 4 safety, or compliance with the specified requirements would result in hardship or unusual

. difficulties without a compensating increase in the level of quality and safety. The attached relief requests are submitted accordingly, i.e., pursuant to 10CFR50.55a(g)(5).

ReliefRequests 4008 and 4010, it may be noted, involve requests for NRC approval to permit IP to utilize alternative requirements, pursuant to 10CFR50.55a(a)(3)

The applicable edition of Section XI of the ASME Code for the first 120-month inservice inspection (ISI) interval at CPS is the 1980 Edition through Winter 1981 Addenda. Inspection activities for the first 120-month ISIinterval are due to be completed during the sixth refueling outage (RF-6) at CPS, currently scheduled to commence on October 13,1996. (RF-6 is the last refueling outage scheduled before expiration of the first ISI intenal.) The planning of work activities for RF-6 is significantly dependent on the NRC's approval of the attached relief requests. IP therefore j

respectfully requests NRCs prompt review and determination regarding approval of the reliefrequests in order that appropriate planning for RF-6 can be completed as soon as

- possible in advance of the outage.

Sincerely yours, f'

^

a Michael W. Lydc Director-Licensing TBE/csm t

Attachments NRC Clinton Licensing Project Manager cc:

NRC Resident Office, V-690 Regional Administrator, Region III, USNRC Illinois Department of Nuclear Safety

to U-602566 Page 1 of 7 ILLINOIS POWER COMPANY Clinton Power Station ASME Section XI Relief Request RELIEF REQUEST 4005 (Revisioti0) i SYSTEM / COMPONENT FOR WHICH RELIEF IS _

Reactor pressure vessel (RPV) head-to-flange weld, Code Class 1. Exammation Category B-A, Item Bl.40, weld

- REQUESTED number CH-C-2.

ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT Addenda, Table IWB-2500-1, Category B-A, Item Bl.40 requires a volumetric examination of the weld and adjacent weld material as defined by Figure IWB-2500-5.

Relief is requested from performing 100% volumetric CODE REQUIRELfENT FROM WHICH RELIEF IS !

exanunation of the exammation volume which is restricted REQUESTED due to scannmg limitations presented by the RPV head j

flange configuration / geometry. It is estimated that 69.6%

of the composite Code examination volume can be examined by ultrasonic examination given that 30.4% is restricted by the configuration / geometry. See attached Figures 1 A through IE.

A portion of the Code-required examination vohune cannot BASIS FORRELIEF be completed due to the RPV head flange configuration / geometry. Figures IA through IE identify the geometrical limitation presented by the RPV head flange with respect to the various ultrasonic examination techniques performed on the head-to-flange weld to date.

Illinois Power Company (IP) proposes to complete

. ALTERNATEEXAMINATIONS -

ultrasonic examination of this weld to the maxunum extent feasible for the required examination volume.

Performance of ultrasonic examination of approximately JUSTIFICATION FOR THE GRANITNG OF 69.6% of the required examination volume prosides RELIEF reasonable assurance of the structural integrity of the entire weld. IP has performed ultrasonic exammation (to the maximum extent feasible) on two-thirds (2S) of the length of this weld based on exammations conducted during RF-2 and RF-4. No unacceptable indications were found. The remammg one-third (la) of the weld length is scheduled to be examined during the next refueling outage (RF-6). IP estimates similar examination coverage (i.e., approximately 69.6%

of tle required exammation volume). It should be noted that during initial plant construction, the entire weld was radiographed and the results were acceptable. The weld was also ultrasonically examined in accordance with the Presenice Inspection Plan, and the results of that examination were also acceptable.

ASME Section V, Article 4, requires that the examination volume (weld and adjacent base material) be scanned by straight and angle beam,45' and 60*, techniques. Article 4 of ASME Section V allows that the examination can be performed from one side of the weld. Due to the flange configuration, ultrasonic examination can only be performed from the head side of the weld. Due to the bend radius on the flange side and the thickness of the flange, and based on presious examination experience with this configuration, only the following examination volunes can be scanned for the techniques listed below (not taking beam spread into account):

to U-602566 Page 2 of 7 IEh._m3lC

% of Examination Volume Examined 0* L-Wave 61.5 %

45' P-Scan 61.5 %

60* P-Scan 61.5 %

45' T-Scan 85.6 %

60* T-Scan 94 %

Composite examination volume = 69.6%

As shown above, at least one technique (60* T-Scan) covers over 90% of the exammation volume, based on previous experience. IP beliews that the actual area exammed using the 45* T-Scan technique was also over 90%, however, if any indication would have been present in the area identified as not eaminal. it would have been difficult to size. For this reason, the examiner identified 85.6% coverage. IP also believes that the actual area previously &aminai using the 60*

T-Scan technique was close to 100% Again, if any indication would have been present in the 6% volume identified as not exammed, it would have been dif5 cult to size. It should be noted, given that the 60 T-Scan technique covers 94% of I

the examination volume, the ASME considers 90% coverage to be acceptable per Code Case N-460. Code Case N460 has been incorporated into Regulatory Guide 1.147.

'Ihe examination limitation imposed by the RPV head-to-flange configuration makes it impractical to perform a complete volumetric exammation (90% or more) of the weld for at least some of the exammation techniques identified above.

Based on the above, however, the structural integrity of the weld can be sufli+a'ly confirmed by completing (in RF-6) the same ultrasonic examination methods and coverage for the remaining length of the weld as done for the first two-thirds of the weld length. IP has thus concluded that completion of the weld examination in this manner provides an acceptable level ofquality and safety.

Reliefis requested for the first ten-year interval at Clinton

. IMPLEMENTATION SCHEDULE Power Station.

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ILLINOIS POWER COMPANY Clinton Power Station ASME Section XI Relief Request RELIEF REQUEST 4006 (Revision 0)

Twenty-nine (29) reactor pressure sessel (RPV) nozzle-to-

. SYSTEM / COMPONINT(S) FOR WHICH RELIEF shell welds, Code Class 1, Exammation Category B-D, IS REQUESTED Item B3.90. See attached Table I for weld numbers.

l ASME Section XI,1980 Edition through Winter 1981 l

CODE REQUIREMENT Addenda, Table IWB-2500-1, Category B-A, Item B3.90 reqmres volumetric exammation of the utld and adjacent l

base material as defined by Figure IWB-2500-7(b).

For each of the identified nozzle-to-shell welds, relief is l

CODE REQUIREMENT FROM WHICH RELIEF IS requested from performing complete volumetric I

REQUESTED examination of the examination vohr which is restricted

~ due to scanning limitations presented by the RPV nozzle conhguration/ geometry. Attached Figures IA through IE depict a typical RPV nozz!c weld, its required examination volume, what portion of the examination volume is restricted by nozzle configuration / geometry, and how that restriction l

affects what percentage of the volume can be examined by each of the various examination techniques (0 L-Wase,45*

l P-Scan,60* T-Scan, etc.). Attached Table 1 identifies the estimates of the composite Code examination volume that can l

be examined for each RPV nozz!c weld.

A portion of the Code-required examination volume cannot

. BASIS FORRELIEF be completed due to the RPV nozzle l

configuration / geometry. Attached Figures IA through IE l

identify the limitations presented by a typical RPV nozzle configuration / geometry for each examination technique employed at Clinton Power Station (CPS).

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Illinois Power Company (IP) proposes to perform l

ALTERNATE EXAMINATIONS ultrasonic examination of these welds to the maximum l

extent feasible for the required examination vohune.

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Performance of ultrasonic examination of approximately JUSTIFICATION FOR THE GRANIING OF 60-70 percent of the required exanunation volume for each RELIEF of the identified welds provides reasonable assurance of the struct ral integrity of each entire weld. IP perfonned ultrasonic examination on 20 of these welds (to the maximum extent feasible) during two presious refueling outages (RF.

2 and RF-4) and found no unacceptable indications. The remairdng 9 welds are scheduled to be examined during the next j

refueling outage. IP estimates similar examination coverage for each of the renuining welds (i.e., approximately 60-70%

i-of the required exammation volume), It should be noted dut during initial plant constmetion, all of these welds were radiographed and the results were acceptable. These welds were also ultrasonically exammed in accordance with the Preservice Inspection Plan, and the results of those examiaations were also acceptable.

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to U-602566 Page 2 of 8 ASME Section V, Article 4, requires that the examination volume (wrid and a4acent base matenal) be scanned by straight and angle beam,45* and 60*, techniques. Article 4 of ASME Section V allows that the exammation can bc performed from one side of the weld. Due to the RPV nozzle configuration, ultrasonic examination can only be performed j

from the shell side. Due to the bend radius and thickness of the nozzle, and based on previous examination experience with this configuration, the following examination volumes can be scanned for a typical nozzle using the techniques listed below (not taking beam spread into account):

1 Techniaue

% of Examination Volume Examined i

0* L Wave 50 %

' 45' P-Scan 60 %

60* P-Scan 72 %

45' T-Scan 80 %

60* T-Scan 88 %

Composite examination volume = 70%

As shown above, at least one technique (60* T-Scan) covers close to 90% of the examination volume. IP believes that the actual area examined using the 45' T-Scan and 60 T-Scan techniques was over 90%, however, if any indication would have been present in the area identified as not examined, it would have been difficult to size. For this reason, the exanuner identified 80% and 88% coverage for the 45* T Scan and 60 T-Scan techniques respectiwly. It should be noted that the ASME, per Code Case N-460, considers 90% coverage as meetmg the Code requirements. Code Case N-460 has been incorporated into Regulatory Guide 1.147.

The examination limitation imposed by the RPV nozzle /shell configuration makes it impractical to perform a complete j

volumetric examination (90% or more) of the nozzle-to-shell welds. However, IP believes that performance of ultrasonic i

exanunation of the welds to the maximum extent feasible (for a composite examination volume of approximately 70%) is i

sufficient for confirming wcld integrity and that such examination therefore provides an acceptable level of quality and l

safety.

Reliefis requested for the first ten-year intenal at CPS.

J IMPLEMENTATION SCHEDULE l

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Page 3 of 8 TABLE 1 l

CODE CATEGORY B-D l

EXAMINATION COVERAGE Estimated Pertentage Of Volume Examined Duc To Weld Number Dgedotion Nozzle Confleuration NIA,B Recirculation Outlet Nozzles 69 N2B,C,D,E,F,G, RecirculationInlet Nozzles 63 i

N2 A, H, J, K RectreulationInlet Nozzles N3A,C Main Steam Nozzles 66 N3B,D Main Steam Nozzles N4A,B,C,D Feedwater Nozzles 66 N5A Core Spray Nozzles 66

. NSB Core Spray Nozzles N6B,C RHR CoolantInjection Nozzles 66 l

N6A RHR CoolantInjection Nozzles N9A Jet Pump Instmment Nozzles 68 I

N9B Jet PumpInstrument Nozzles NIO CRD Return Nozzles 68 N16 VibrationInstrument Nonles 66 These weld examinati'>ns will be completed during the next refueling outage (RF-6). IP expects the percentage of volume examined for each weld to be between 60 and 70 percent.

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Figure 1A A Typical RPV Nozzle to Shell Weld

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[_] Code cross-sectional coverage 51.4 sq in Veld netal L-wave examination coverage 25.6 sq in I

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Figure 1B A Typical RPV Nozzle to Shell Weld

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[-] Code cross-sectional coverage 51.4 sq in y 45* T-scan exantnation cowerage 41.1 sq in

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Figure 1C A Typical RPV Nozzle to Shell Weld

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[_~] Code cross-sectional coverage 51.4 sq in y 45' P-scan examination coverage 30.9 sq in

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Figure 1E A. Typical RPV. Nozzle to Shell Weld

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ILLINOIS POWER COMPANY Clinton Power Station ASME Section XI Relief Request RELIEF REQUEST 4007 (Revision 0) l l

Feedwater system pipe-to-valve weld 1-FW-2-5, Code l

1 SYSTEM / COMPONENT FOR WHICH RFI JFF IS Class 1, Exammation Category B4, Item B9.11.

REOUESTED ~-

ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT Addenda, Table IWB-2500-1, Examination Category B-J, Item B9.ll, Note 1(b), requires an outside surface exammation of the wetJ and adjacent base metal and a i

volumetric exammation of the weld and adjacent base n:etl (lower one-third volu neus defined by Figure IWB-2500-8..

] Relief is reqv.sted from selecting weld 1-FW-2-5 for CODE REQUIREMENT FROM WHICH RELIEF IS pedormance ; f surface and volumetric examinations i

REQUESTED deg the fin t ten-year interval.

The subject weld is a pipe-to-valve weld on one of the two, BASIS FORRELIEF main 18-inch feedwater lines inside the drywell. The associated feedwater pipe (the "B" line) is supponed by a large component support (guide support) that supports the l

pipe at the weld location. In order to perform the Code-required exammation, the guide support would have to bc l

disassembled to gain access to this weld. A drawing of this large and complicated guide support is attached (Figure 1).

l Illinois Power Company (IP) proposes, in lieu of selecting l

ALTERNATEEXAMINATIONS this weld for examination, selecting a difTerent weld in the Feedwater system for examination.

I In order to perform the Code-required exammation, guide JUSTIFICATION FOR 'ITIE GRANTING OF suppon IFW01014G would have to be disassembled to RELIEF gain access to this weld. Remosul and reassembly of the j

guide support is a tedious process. For example, a l

temporary support would have to be installed prior to removing the guide support from the piping. Also, some of the l

replacement pans that may be required, such as the associated Lubron plates, are not readily asailable. Based on presious l

experience at Clinton Power Station (CPS), i e., during RF-2 when the sister guide support (for the "A" feedwater line) was removed for examinaticn of weld 1-FW-1-5, hundreds of man-hours and several man-rem of exposure must be expended to disassemble and reassemble the guide support. Finally, notwithstanding the cost and effort, IP belieses it is imprudent to disassemble this properly functioning guide suppon in light of IP's conclusion that a sufficient number of welds have been examined (or are scheduled to be examined during the next refueling outage) as further discussed below.

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Page 2 of 7 l

The reason for selecting weld 1-FW-2-5 for examination is that, based on the associated system stress report, this weld is subject to stress that exceeds the stress criteria identified in Note 1(b) of Table IWB-2500-1 for Category B-L llowever, there are two more Feedwater system welds subject to stress !cvels that exceed the Note 1(b) stress criteria, and which are l

in the same Feedwater loop ("B*) as the subject weld. Dese two welds,1 FW-2-3-8 and 1-FW-2-3-9, are scheduled for l

cxamination in the next outage In addition, there are three welds in Feedwater loop "A" at smular locations, for which l

apphed stresses also exceed the Note 1(b) stress criteria. All of those welds,1-FW-t-5 (mentioned presiously),1-FW-1 '

8 and 1-FW-1-3-9, have been examined, and the results were acceptable. Dus, by the conclusion of the next outage, five of the six welds to which Note 1(b) applies will have been exarmned. (The relative locations of these welds in the l

fealwater piping is shown on the attached isometric drawings, i.e., Figures I and 2 for feedwater loop "A," and Figures 2 and 3 for feedwater loop "B.") In addition, for all welds required to be e>=in~i in the feedwater system (loops "A" and "B"), approximately 27% of all applicable welds have been exammed in the last fhe (5) outages, relative to the Code requirement of 25% over the ten-year interval. These welds were found to have no un=~eble indications. Based upon these examinations, IP believes the intent of the Code requirements for verifying the structural integrity of the feedwater system has been met. On that basis, selecting another weld in the feedwater system for exammation instead of weld

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l-FW-2-5 will still proside an acceptable level of quality or safety for the feedwater system.

Reliefis requested for the first ten-year intenal at CPS.

IMPLEMENTATION SCHEDULE 1.

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I to U-602566

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Page 1 of 3 i

I ILLINOIS POWER COMPANY Clinton Power Station ASME Secdon XI Relief Request RELIEF REQUEST 4008 (Revision 0)

)

Fv=inMion Category B-K-1, Items B10.10 and B10.20

. SYSTEM / COMPONENT (s) FOR WHICH RFUFF (integral attachments for piping and pumps, respectively),

'IS REQUESTED Code Class 1: (1) Piping Attachments / Lugs: Weld Numbers 1-MS-A-7PR-WA,1-MS-B-8-PR-WA,1-MS-C-8-PR-WA and 1-MS-D-7-PR-WA (for Main Steam system (MS) guide supports), and IRR-A-PR 1-WA (for Reactor l

Rectreulation system (RR) variable support); (2) Pump Attachments / Lugs: (Unnumbered) Welds for shock suppressor lugs attached to Reactor Recirculation Pump B bowl.

ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT Addenda, requires surface examination of welds for integral attachments during the first ten-year interval.

Relief for Item B10.10 is requested from performing a full CODE REQUIREMENT FROM WHICH RELIEF IS surface exammation of the identified welds for the MS and REQUESTED RR piping attachments / lugs as access to each of the i

exanunation areas is restricted due to the associated l

component supports. For Item B10.20, relief is requested from performing surface examination of any of the shock l

suppressor lug welds for RR Pump B.

l For item B10.10: Performing the required surface l

BASIS FORRELIEF examination of the entire exammation area for each of the welds associated with integral attachments or lugs on each of the main steam lines and on the RR Loop "A" suction line requires the removal / disassembly of four guide supports (one for each steam line) and one variable support (for the RR loop "A" suction line) in order to gain access to the required exammation areas. For Item B10.20: In order to perform the required surfhce c:. amination of the shock suppressor lugs on Reactor Recirculation Pump B the entire insulation surmunding the pump bowl, which is located in a high radiation area, would have to be removed.

For Item B10.10: Perform a surface exammation of each of ALTERNATE EXAMINATIONS the weld areas to the maximum extent feasible without removal or disassembly of the associated component supports. Table 1 identifies the exammation coverage l

feasible based on exammations that have been completed to date and on what is projected to be done for the remaining welds to be examined in the next scheduled refueling outage. For Item B10.20: Utilize ASME Code Case N-509 issued by the ASME on Nov.mber 25,1992, which allows examination ofintegral attachments on a sample basis.

For Item B10.10 - Main steam line attachments: Based on JUSTIFICATION -FOR THE GRANTING 'OF the size, design and installation of tir Main Steam guide RELIEF supports, disassembly and rem sly of these guide supports (one for each steam unw would be a complicated and tedious process, imching a difIicult procedure to implement. (A drawing of one of the main steamline guide supports is attached as Figure 1.) Funher, although none of the guide supports were removed for any of the attaclunent weld examinations performed to da!c for the main steam lines, based on the presious disassembly of a similar guide l

suppon in another system, Illinois Power Company (IP) believes that completion of the process would require replacing i

i

~

to U-602566 Page 2 of 3 the Lubron plates for the supports, which are difficult to procure. The intent of the ASME Section XI exammation is to prmide assurance of structural integrity rather than require disassembly and reassembly of properly functioning components / supports, possibly adversely affecting their operability. Disassembly of a property functioning guide support is thus notjudged to be prudent. In addition, due to the extensive effort required, disassembly and reassembly of the main steam line guide supports would likely extend the duration of the outage and result in on-'y radiation exposure For Item B10.10 - Reactor recirculation piping attachment: The uriable support for the associated Reactor Recirculation system pipe is a load <arrying support. In order to remove this support, a temporary support would be required to be installed. The variable support is located 13 feet above the floor elevation and is in a high radiation area. Several man.

hours would have to be expended to erect scaffolding, remove insulation, install a knyoisry support and transfer the piping load prior to remming the variable support. Further effort would then be required for system restoration. The total dose that would be incurred is estimated to be five (5) man-rem. Examining 65 percent of the weld area (to facilitate examination of the obstructed weld area for the attachment / lug on the associated reactor rectrculation line) should be -

sufficient to establish the integrity of the lug weld without significantly reducmg the safety margin prmided by such verification.

For Item B10.20 - Shock suppresser lugs on RR Pump B: The ASME published Code Case N-509," Alternative Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded AttachmentsSection XI, Di3ision 1", on November 25,1992. This Code Case requires examination of a sample of 10% of the welded attachments on pumps in lieu of 100%

examination. IP has performed examination of all the welded attachments en RR Pump A (which is essentially identical to RR Pump B). This is equivalent to a 50% sample of the Reactor Recirculation System. Utilizing Code Case N-509 will reduce the undue burden on IP without reducing the safety margin prmided by verification of wrld integrity.

The requested relief and proposed use of an alternate IMPLEMENTATION SCHEDULE examination approach are for the first ten-year intenal at CPS.

TABLE 1 l

CODE CATEGORY B-K-1 EXAMINATION COVERAGE Estimated Perrentage of Area Examined Due To Component Support (s)

Weld Number

System Description

Interference IMS-A 7PR-WA Main Steam 25 1MS-B-8PR-WA Main Steam IMS-C-8PR-WA Main Steam 27 IMS-D-7PR-WA Main Steam 27 1RR-A-PR-1-WA Reactor Recirculation 65

• This item is scheduled for examination during the next refueling outage. IP expects the percer me of area examined to be 25 to 27 percent.

I l

to U-602566 Page 3 of 3 1

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to U-602566 Page l of 2 ILLINOIS POWER COMPANY Clinton Power Station AShE Section XI Relief Request RELIEF REQUEST 4009 (Revision 0)

Examination Category B-L-2 & B-M-2, Item B12.20 &

- SYSTEM / COMPONENT (s) FOR WHICH RELIEF B12.50, Code Class 1: 1 Pump, Reactor Recirculation IS REQUESTED Pump A (IB33-C001A) or Reactor Recundation Pump B (IB33-C001B), and 11 Valves, IB21-F0llA, IB33-F023A, IB33-F060A, IB33-F067A, IE12-F008, IE12-F010, IE12-F039A, IE12-F041 A, IE21-F007, IE22-F036 and IG33-F102 ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT '

Addenda, requires sisual (VT-3) exammation of Class 1 pump casing internal surfaces (Category B-L-2) and Class 1 vahr body internal surfaces (Category B-M-2) during the first ten-year inspection interval.

Relief is requested from performing the Code-required CODE REQUIREMENT FROM WHICH RELIEF IS visual (VT-3) exanunation of pump and vahr internal REQUES'IED surfaces on the items identified above.

In order to perform the required sisual (VT-3)

BASIS FORRELIEF examinations, Clinton Power Station personnel would have to dmwmble the afrected components. Hundreds of man-hours and several man-rem of exposure would be expended for the erection / dismantling of scaffolding, removal / reinstallation of insulation, and disassembly / reassembly of the pump and vahes only to perform these VT-3 examinations. Illinois Power Company (IP) has determined that the hardship imuhtd in performing the VT-3 examinations is umvarranted and that it is impnident to disassemble properly functioning pumps or vahesjust for the sake of such exanunations.

In lieu of the requirements of ASME Section XI,1980 ALTERNATE EXAMINATIONS Edition through Winter 1981 Addenda, for Categories B-L-2 and B-M-2, IP proposes to utilize AShE Section XI, 1989 Edition, which has been incorporated into 10 CFR50.55a (b).

It is impnident to disassemble properly functioning pumps JUSTIFICATION FORTHE GRANTING OF or vahts only to perform VT-3 sisual examinations of RELEF internal surfaces, nor is it good ALARA practice to incur exposure of several man-rem only for such examinations.

The 1989 Edition of AShE Section XI reflects a change to the requirements for VT-3 examinations such that VF-3 internal-surface examinations of pumps and vah es are only required when pumps and valves are being disassembled for maintenance. It would be an unnecessary burden on Illinois Power to disassemble / reassemble the identified pump and vahes solely to comply with the 1980 Edition (through Winter 1981 Addenda) of ASME Section XI in light of the fact that the NRC has incorporated a newer (1989) Edition into 10CFR50.55a(b) which only requires the VT-3 visual examinations to be performed when the pumps or vahrs are being disassembled for maintenance. Resision of the Code and its acceptance by the NRC and the industry confirms that performance of VT-3 examiriations during pump or vahr disassembly for maintenance is sufficient for examining internal surfaces for component integrity relative to disassembling

to U-602566 Page 2 of 2 l

j components just to perform such examinations. Considering the potential effects of unnmery disassembly and the.

l radiation exposure invohed in performing such work and in performing the VT-3 exanunations themsches on a periodic l

basis, there does not appear to be any overall safety benefit to be gamed by performing VT-3 examinations in accordance with the Edition of the Code to which IP is currently committed (for the first ten-year inten al).

Reliefis requested for the first ten-year inten~al at CPS.

1 IMPLEMENTATION SCHEDULE l

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I to U-602566 Page 1 of 4 ILLINOIS POWER COMPANY Clinton Power Station ASME Section XI Relief Request RELIEF REQUEST 4010 (Revision 0)

' examination Category B-A, item Bl.ll and Bl.12 SYSTEhV COMPONENT (s) FOR.WHICH RELIEF "

Code Class 1,

Reactor Pressure Vcssel (RPV)

IS REQUESTED circumferential welds RPV-Cl and RPV-C2, and vertical welds RPV-VI A, RPV-V1B, RPV-V2A, RPV-V2B and RPV-V2C.

ASME Section XI,1980 Edition through Winter 1981 l

CODE REQUIREMENT Addenda, Table IWB-2500-1, Examination Category B-A, Item Bl.ll & Bl.12, requires 100% volumetric examinations of RPV circumferential and vertical welds during the first ten-year inspection interval as defined by Figure IWB-2500-1 and IWB-2500-2.

Relief is requested from performing the Code-required CODE REQUIREMENT FROM WHICH RELIEF IS ;

volumetric examination of the RPV circumferential and

-REQUESTED vertical welds identified above.

The subject welds are located behind the biological BASIS FORRELIEF shield wall (surrounding the RPV) where the radiation levels are extremely high. Even after nozzle flushing, Illinois Power Company (IP) estimates a personnel exposure of 20 man-rem to perform the required inspections. In addition, IP estimates the cost of performing examinations of these welds to be approximately $400,000.

On September 28,1995, the BWR Vessel and Internals Project (BWRVIP) submitted a topical report to the NRC, entitled "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", EPRI TR-105697, which recommends alternative examination criteria for Category B A, Item Bl.Il and Bl.12 welds. The conclusions and recommendations in this report are consistent with 1P's request for relief to reduce the scope of required examinations for the applicable welds.

IP proposes to adopt an inspection program of reduced AL'IERNATE EXAMINATIONS scope for the number of RPV circumferential and vertical welds to be examined, in accordance with the recommendations of EPRI TR-105697.

Clinton Power Station (CPS) has four (4)

JUSTIFICATIONFORTHE GRANIING OF circumferential welds per item No. Bl.11 (RPV-Cl RELIEF through RPV-C4) and 11 vertical welds per Item No.

Bl.12 (RPV-VIA through RPV-V4C). During plant construction, as required by AShE Section III, all of these welds were examined by surface and radiographic examination methods, and accepted as free of any manufacturing flaws. As part of the Presenice Examination Plan, CPS also performed ultrasonic examination on these welds, and they were accepted for placing the RPV in senice. Further, a hydrostatic test at over 1500 psi was satisfactorily performed in accordance with AShE Section

l o

l l

to U-602566 Page 2 of 4 Ill requirements. With respect to assembly of the RPV itself, the vessel was built under rigorous requirements for fabrication and examination, and post weld heat treatment during fabrication assured that shell weld and cladding I

residual stresses were minimized. Thesc_ controls and actions, along with all of the above-noted surface, radiographic, and ultrasonic examinations, greatly reduced the likelihood of large flaws existing in the shell weld, j

. especially at the vessel surfaces, and thus contributed to providing assurance of RPV integrity.

During the first ten-year inspection interval (i.e., during refueling outages RF-2 and RF-4), the following welds were ultrasonically examined and accepted for continued operation:

1 Circumferential Welds ner Item No. Bl.11 Vertical Welds ocr Item No. Bl.12 RPV-C3 RPV-V3A, V3B, V3C RPV-C4 RPV-V4A, V4B, V4C i

- The above welds comprise 50% of the circumferential welds for Item Bl.11 and almost 55% of the vertical welds l

for Item Bl.12. The examinations performed for these welds (as well as the ultrasonic examinations performed as part of Preservice Inspection Plan) were performed in accordance with Reg Guide 1.150. In addition, other welds or portions of other welds on the RPV were examined by ultrasonic method during RF-2 and RF-4 and accepted for l

continued service. These include 50% of the shell-to-flange weld, over 60% of the top-head-to-flange weld, some top-head welds, portions of the bottom-head-to-support-skirt weld, and several nozzle-to-shell welds.

A combination of BWR operating characteristics and good material properties makes CPS inherently flaw tolerant -

- during operation. The operating characteristics of a BWR like CPS preclude extreme brittle fracture combinations of pressure and temperature (due to saturated conditions for operating BWRs), such that the periodic RPV pressure i

test is the most limiting condition. These favorable operating characteristics are acknowledged in EPRI TR-l 105697. Further, the EPRI report discusses a survey of vessel inspection information that was conducted for BWRs. The following results were reported:

l For the 24 BWRs included in the survey, a cumulative total RPV weld length of 14,565 ft is currently required to be examined,: Of the total 14,565 ft of weld,4770 A has undergone full Code examination.

Another 487 ft of weld has undergone partial Code examination. The total length of weld examined to 1

date therefore equals $257 ft, or 36% of the total possible weld length. In the 5257 ft of weld examined to date, only 16 indications were found that did not meet the acceptance criteria of ASME Section XI, IWB-3500. All 16 of those indications were subsurface flaws that were shown to be acceptable by meeting the criteria ofIWB-3600.

It is clear from the survey results that a substantial amount of examination has been performed to verify the I

integrity of BWR Vessels, and that only a negligible number of indications have been detected as a result. Per EPRI TR-105697, it has been concluded that the Code-required volumetric inspection scope for BWR RPV shell welds (Category B A, Item Numbers Bl.ll & Bl.12) during each ten-year interval may be replaced with the following volumetric inspection requirements, without a significant impact on BWR integrity:

1.

Inspect 50% of longitudinal seam welds (Category B-A, item No. Bl.12) in the reactor vessel shell. This requirement maybe satisfied by inspecting 30% of the total weld length of this category of weld, using any combination of the total number of welds or percentage of each weld inspected to achieve this overall percentage.

2.

Inspect 0% of the circumferential seam welds (Categosy B-A, Item No. Bl.11) in the reactor vessel shell, l

3.

Inspection procedures used for these examinations shall be qualified such that flaws relevant to i

vessel integrity shall be reliably detected and sized. Personnel implementing these procedures shall be qualified in the use of these procedures.

i b

o l

to U-602566

{

Page 3 of 4 Attached Table 1, which is based on Table 9-1 in EPRI TR-105697, identifies the calculated probabilities of RPV i

failures and leakage per 40 vessel years due to postulated crack development and growth under each of the two evaluated inspection programs, i.e., the Code-required program versus the BWRVIP/EPRI-recommended program.

d Based on the resultant RPV failure probability of 1.151x10 /40 vessel years (for longitudinal and circumferential welds) as obtained for the BWRVIP/EPRI proposed inspect. ion program, the RPV failure probability under the proposed program would still be more than two orders of magnitude less than the NRC safety goal for reactor 4

j vessel failure of1x10 per vessel year or 4x10'8 in 40 vessel years.

]

As previously noted, IP has ultrasonically examined 50% of the RPV circumferential welds and almost 55% of the RPV longitudinal welds, based on examinations completed during RF-2 and RF-4. This is significantly more than the number of welds required per EPRI TR-105697 wherein an examination of 0% and 50% of the circumferential and longitudinal welds, respectively, was shown to still support the safety goal for failure probability of the RPV.

Considering the estimated cost of examining the remaining welds, both in dollars and in person-rem, the hardship or cost of performing the additional inspections is not justified relative to the marginal safety benefit estimated to be gained by doing the additional inspections.

IMPLEMENTATION SCHEDULE Reliefis requested for the first ten-year interval at CPS.

)

Attaclunent 6 to U-602566 Page 4 of 4 TABLE 1 EFFECT OF ISI PROGRAM PROPOSED PER EPRI TR-105697 Probability of Failure Probability of Leakage Per 40 Vessel Years Per 40 Vessel Years Code Proposed Code Proposed Requirements' Program Requirements' Program:

2 Long. Welds 5.68x 10*

1.15x 10 1.45x 10 1.34x 10-5 4

4 (Irradiated)

Long. Welds 1.07x10-'

1.13x10-'

5.0lx10 2.39x10-5 4

(Unirradiated)

Circumferential 1.26x10" 1.10x10" 1.94x10 22 2.12x10 22 Welds (All) 4 4

4 Totals 5.69x10 1.151x10 6.46x10 3.73x10 5

1. 90% Inspection of all Longitudinal and Circumferential Welds 2 50% Inspection of Longitudinal Welds; No Inspection of Circumferential Welds I

l l

I

to U-602566

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Page 1 of 2 ILLINOIS POWER COMPANY Clinton Power Station i

ASME Section XI Relief Request RELIEF REQUEST 4011 (Revision 0)

Examination Category C-A, item C1.20, Code Class 2 -

SYSTEM / COMPONENT FOR WHICH BMMIS head-to-vessel weld on Residual Heat Removal (RHR) Heat REQUESTED Exchanger A, weld numberHEA-1.

ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT Addenda, requires ultrasonic' examination of the head-to-vessel weld on RHR Heat Exchanger A during the first ten-year interval.

Reliefis requested from performing ultrasonic examination CODE REQUIREMENTS FROM WHICH RELIEF of approximately 3% of the examination volume since only IS REQUESTED approximately 87% of the required examination volume can be examined by ultrasonic examination.

Four permanently welded / installed lifting lugs interfere BASIS FORRELIEF with the ultrasonic examination. To perform examination of the entire volume, the lifting lugs wodd have to be removed by grinding. A drawing of a representative lifting lug is attached as Figure 1.

Perform the ultmsonic examination of the accessible area to ALTERNATE EXAMINATION the maximum extent feasible without remosing the permanently installed lifting lugs.

ASME Code Case N460 has been incorporated into JUSTIFICATION FOR' THE-GRANTING OF -

Regulatory Guide 1.147, Rev.11 dated October,1994. This RELIEF Code Case allows a reduction of the examination area of up to 10% Another 3% reduction in the examination area for this weld would not significantly impact the effectiveness of the examination and should thus not compromise safety. To examine the required 3% additional exmnination area, Clinton Power Station (CPS) personnel would have to remove the four lugs by grinding. This would require erecting very brge scaffolding all around the heat exchanger and remosing an extensive amount ofinsulation. Further, this RHR heat exchanger is located in a high radiation area. All of this effort to support and perform examination of the additional weld area would thus require Illinois Power Company to expend many unnecessary man-hours and incur a significant amount of exposure (estimated to be 3-4 man-rem).

Reliefis requested for the first ten-year inten al at CPS IMPLEMENTATION SCHEDULE

• w GEN ER AL h ELECTRIC PR J cT:

SYSTEM: N 'i REPORT NO. : 635-ossw SHEET NO.

• b WELD NO.: M E A-1 UNIT NO.:

I

= n+:

RHR A

(

0 0*, W, I 80*, 210~ A2.

HEAT Evtaniceu

,/,y 4 /

l lie 4iTATico I 00 EbCR 6tDE OF LivTiNG lug 6 G 0*,90*, t 80*,210' At.

._-5.15--.

.-i.cd w6nS WELO k WELD

• REA-l WfLD* AEA-l IN l

"i i

w.,

a

?

O' BY:

b DATE:

' /3/3

LEVEL:

f PAGE u/' 0F "k Figure 1 4

m_ _ _

____._.__.2m

___ mum---

0 s

4 to U-602566 Page 1 of 3 ILLINOIS POWER COMPANY Clinton Power Station ASME Section XI Relief Request l

RELIEF REQUEST 4012 (Revision 0)

Examination Category C-C, Item C3.30 - Integral

. SYSTEM / COMPONENT FOR WHICH RELIEF IS ;

attachments on Class 2 pump casing, i.e., on the Reactor REQUESTED Core Isolation Cooling (RCIC) pump, weld number RCIC-1 A(14).

ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT Addenda, requires surface enmination of integral l

attachments during the first ten-year interval.

Reliefis requested from performing surface examination of CODE REQUIREMENfS FROM WHICH RELIEF the integral attachment welds for approximately 6% of the IS REQUESTED exanunation area since only 84% of the examination area can be examined by surface examination method.

There are four identical lugs (integral attachments) welded BASIS FORRELIEF to the RCIC punp casing which are utilized to mount the pump to its pedestal. Once the pump is installed, the lower side of the weld (on each of the four lugs) is not accessible for examination. In order to perform the exandnation on the lower side, the pump itself would have to be removed from the pedestal Attached Figure 1 is a drawing and photograph indicating the positions cf the pump lugs on the RCIC pump casing. Figure 2 shows what was examined for each of the !ug welds for the surface examinations performed to date.

~

Perform the surface examination of the accessible area to

' ALTERNATEEXAMINATIONS the maximum extent feasible without remosing the pump from its pedestal.

To gain access to the lower side of the lugs, the pump JUSTIFICATIONFORTHE GRANTING OF would have to be removed from the pedestal Apart from RELIEF performing the examination itself, this would require disconnecting the piping from the pump, re-installing the pump back on the pedestal, reconnecting the piping, and verifying that everything is installed properly and meets design requirements. This process would require Illinois Power Company (IP) to expend sestral num-hours and would result in unnecessary radiation exposure (estunated to be approximately 2 man-rem). NRC has incorporated Code Case N-160,

" Alternative Examination Coverage for Class 1 and 2 welds", into Regulatory Guide 1.147, Rev. I1 dated October,1994.

This Code Case allows a reduction in the examination coverage up to 10%. In the case of this weld, RCIC-1 A (1-4), IP l

cannot perform 16% of the examination area due to an interference with the pedestal A reduction of another 6% in the l

exanunation coverage will not significantly reduce the margin of safety prosided by performing the surface examination for verifying weld integrity.

Reliefis requested for the first ten-year intenal at Clinton IMPLEMENTATION SCHEDULE Power Station.

l l

. ~... _ _... _ -

...._.m...._

~.._...__.._..._..----...___.....___.-.m_____

)

SIZE:

f ACCESS INFORMATION KIO8 scH"-

I FIG RI- 08-3 AT.:

OlSCHARGE FLANGE

(

ANSI 9004 SG ASTM SPEC:

i 3

5 NOM WALL:

CAL BLK:

l SC:60 BY PASS f

4 RF f

1(-1500 PUMP LUGS (2 LU6S EACH SIOE) t

/

(I A(1-4D l

l l

i SUCTION FLANGE l

ANSI 3OO2* RF l

y h

RIO6 s

(D i

THIS DRAWING HAS BEEN REDRAWN l

~

FROM FORMER DRAWING NUMBER B-78 7R!

EEl n

FIGURE NO.:

RI 5 ol!

~

REVISION NO.:

0 DATED: 2-3-92

[E; P&lD NO.:

b

?;'

REFERENCE DWG. NO.:

8i N

Figure 1

- GR

.m,

.m.

G-RCIC-1 A ( 9 -G) f l

TCrTAL EXAMtNABLE AREA TOTS. AR!A EXAMINED LENGTH WCTH LENGTM WOTH A1 95 x 0.5 =

4.75 At 9.5 x

0.5 =

4.75 4.25 A2 8.5 x

0.5 =

4.25 l

A2 8.5 x

0.5

=

A3 7.5 x 0.5 =

3.7S A3 7.5 x 0.5 =

3.75 TOTAL 12.75 TOTAL 12.75 2

8 WBeeM 11.5 81 23 x 0.5 =

115 B1 23 x 05 =

10 82 20 x 0.5 =

to i

$3 WeuMeed B2 20 x 0.5

=

h 8.5 53 17 x 0.5 =

8.5 83 17 x 05

=

9 9

t/2* LesMetal TOTAL 30 TOTAL 30 A

C 4.75 C1 9.5 x 0.5 =

4.75 Img C1 9.5 x

0.5

=

C2 8.5 x

0.5 =

4E C2 8.5 x 0.5 =

43 C3 7.5 x 05 =

3.75 C3 7.5 x 0.5 =

3.75 i

TOTAL 12.75 TOTAL 12.75 CLAMP AREA D1 8 x 05 =

4 D1 0 x 0.5 =

0 3.5 D2 0 x 0.5 =

0 I

D2 7 x 05

=

D

=

3 D3 0 x 0.5 =

0 D3 8 x 05 TCTAL 105 TOTAL 0

TOTAL AREA = ANx4 LUGS TOTAL AREA = A+B+C+0 x 4 LUGS wTAL mA=

2M TOTE MA =

222 g%

nn (D (D TOTAL CODE COVERAGE =

84.09 %

u O

- =,

n W

Figure 2 g

n 0

C O

63

• D

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o.

e to U-602566 I

Page1ofI ILLINOIS POWER COMPANY Clinton Power Station l

ASME Section XI Relief Request RELIEF REQUEST 4013 (Revision 0) l Examination Category C-G, Item C6.10 - Pump casing l

o SYSTEM / COMPONENT FOR WHICH RELIEF IS -

weld on Class 2 pump, i.e., on the "A" Residual Heat

REQUESTED Removal (RHR) pump, weld nunber RHR-A-2.

(

ASME Section XI,1980 Edition through Winter 1981 CODE REQUIREMENT Addenda, requires surface examination of the pump casing j

welds during the first ten-year interval.

Reliefis requested from performing surface exammation of CODE REQUIREMENTS FROM WHICH RELIEF approximately 3% of the examination volume for this weld IS REQUESTED since only approximately 87% of the enmination volume can be examined by surface examination method.

A permanently installed instrument line interferes with the l

BASIS FORRELIEF surface examination of this weld. To perform a full Code-l required examination of this area, the mstrument line would have to be cut out to gain access-i Perform the surface examination of the accescible area to ALTERNATEEXAMINATION the maximum extent feasible without removing the permanently installed instrument line.

Code Case N-460 has been incorporated into Regulatory l

JUSTIFICATION FOR THE GRANTING OF Guide 1.147, Rev.11 dated October,1994. This Code l

RELIEF -

Case allows a reduction in the examination area of up to l

10%. Another 3% reduction in the enmination area l

would not significantly reduce the effectiveness of the examination for verifying weld integrity. To examine the Code-required 3% additional examination area, Clinton Power Station (CPS) personnel would have to cut out the noted instrument line and weld it back. Funher, the effort required would result in unnecessary radiation exposure to plant personnel. Relative to the little or negligible safety benefit gained by examining the obstructed weld area, removal and repair of the instnunent line, along with the radiation exposure incurred, is notjustified.

Reliefis requested for the first ten-year inten'al at CPS.

IMPLEMENTATION SCHEDULE

,